Text

Redacted - Susquehanna Steam Electric Station, Units 1 & 2, Revision 68 to Final Safety Analysis Report, Questions 005.1 to 040.99
	ML18124A200
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Site:	Susquehanna
Issue date:	10/16/2017
From:	; Susquehanna, Talen Energy
To:	; Office of Nuclear Reactor Regulation
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ML17331A584	List: ML17331A532; ML17331A533; ML17331A534; ML17331A547; ML17331A551; ML17331A553; ML17331A554; ML17331A564; ML18116A085; ML18116A111; ML18116A538; ML18117A090; ML18122A329; ML18122A338; ML18122A369; ML18123A242; ML18123A330; ML18123A485; ML18123A510; ML18124A094; ML18124A109; ML18124A125; ML18124A152; ML18124A200; ML18124A204; ML18298A237; PLA-7636, Redacted - Susquehanna Steam Electric Station, Units 1 & 2, Revision 68 to Final Safety Analysis Report, Chapter 3, Section 3.9 to 3.14, Conformance with NRC General Design Criteria. Part 2 of 2;
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SSES-FSAR Rev. 46, 06/93 005.1-1 QUESTION 005.1 The statement in Section 5.2.1.1 of the FSAR with regard to your compliance with

10 CFR Part 50, Section 50.55a, Codes and Standards Rule, is incorrect as a number of Quality Group A components within the reactor coolant pressure boundary are not in conformance with the applicable ASME Boiler and Pressure Vessel code and addenda as required by the rule.

In Amendment 13 to the Susquehanna Steam Electric Stations' FSAR and in your letter ER 100450, File 040-2, received by the Staff on March 1, 1974, you provided an analysis of anticipated deviations from the codes and standards rule requirements set forth in the provisions of Section 50.55a, 10 CFR Part 50, based on a Construction Permit Date of November 2, 1973, for the Susquehanna reactor pressure vessels, reactor recirculation piping, reactor recirculation system pumps, main steam line isolation valves, and main steam safety/relief valves. Based on this information and on certain additional commitments relative to the reactor pressure vessels, the AEC in a letter dated June 20, 1974, in accordance with paragraph 50.55a (a)(2)(ii), granted approval for relief from the rule for these components and acceptance of the ASME Section III Code and Addenda specified in Amendment 13 to the FSAR and letter ER 100450, File 040-

2. Revise Section 5.2.1.1. of the FSAR to correctly reflect the status of each Quality Group A component within the reactor coolant pressure boundary.

RESPONSE:

For response see Subsection 5.2.1.1 and Table 5.2-

10.

SSES-FSAR Rev. 46, 06/93 005.2-1 QUESTION 005.2 In Table 3.2-1 of the FSAR identify the applicable principal construction codes and standards in those cases where this information is now missing throughout the table.

RESPONSE:

Table 3.2-1 has been revised to provide the requested information.

SSES-FSAR Rev. 51, 02/97 005.3-1 QUESTION 005.3 The B31.1 component code identified in Table 3.2-1 of the FSAR for the diesel lube oil system piping and valves is inconsistent with the Quality Group C (Safety Class 3) classification for these components. The diesel generator lubrication system piping is also identified in Section 9.5.7.1 of the FSAR as designed in accordance with ASME Section III, Class 3. Resolve this inconsistency and revise the FSAR as appropriate.

RESPONSE:

Section 9.5.7.1 of the FSAR has been revised to resolve this inconsistency.

SSES-FSAR Rev. 46, 06/93 005.4-1 QUESTION 005.4 The quality group (safety class) classification, seismic classification, component code and quality assurance requirements for the components of the Emergency Service Water System have been omitted from Table 3.2-1. Revise Table 3.2-1 to include this information.

RESPONSE:

Table 3.2-1 of the FSAR has been revised to include this information.

SSES-FSAR Rev. 46, 06/93 005.5-1 QUESTION 005.5 The quality group (safety class) classification, seismic classification, component code and quality assurance requirements for the spray pond system piping has been omitted from Table 3.2-1. Revise Table 3.2-1 to include this information.

RESPONSE:

Table 3.2-1 of the FSAR has been revised to include this information.

SSES-FSAR Rev. 46, 06/93 005.6-1 QUESTION 005.6 Verify that all components within the reactor coolant pressure boundary as defined in

10 CFR Part 50.2(v) are classified Quality Group A in compliance with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50, or as a minimum, are classified Quality Group B if the components meet the exclusion requirements of the rule.

RESPONSE:

Section 3.1.2.2.5 of the FSAR has been revised to resolve this concern.

SSES-FSAR Rev. 46, 06/93 010.1-1 QUESTION 010.1 The criteria for your high energy and moderate energy line analysis

in the FSAR is in accordance with Branch Technical Position APCSB

3-1, "Protection Against Piping Failures in Fluid Systems Outside

Containment." However, other than Tables 3.6-2 and 3.6-3, the

results of your analyses and the environmental effects regulating

from high energy line breaks and leakage cracks have not been

provided. Provide these analyses and results for each of the

assumed breaks or leakage cracks at their postulated locations.

RESPONSE:

The requested information has been provided in Appendix 3.6A.

SSES-FSAR Rev. 50, 07/96 010.2-1 QUESTION 010.2 We require that the compartment between the containment and the reactor building which houses the main steam lines and feedwater lines and the isolation valves for those lines, be designed to consider the environmental effects (pressure, temperature, humidity) and potential flooding consequences from an assumed crack, equivalent to the flow area of a single ended pipe rupture in these lines. We require that essential equipment located within the compartment, including the main steam isolation and feedwater valves and their operators be capable of operating in the environment resulting from the above crack. We also will require that if this assumed crack could cause the structural failure of this compartment, then the failure should not jeopardize the safe shutdown of the plant. In addition, we require that the remaining portion of the pipe in the tunnel between the reactor building and the turbine building meet the guidelines of Branch Technical Position APCSB 3-

1. We require that you submit a subcompartment pressure analysis to confirm that the design of both areas conforms to our position as outlined above.

We request that you evaluate the design against this staff position, and advise us as to the outcome of your review, including any design changes which may be required. The evaluation should include a verification that the methods used to calculate the pressure build-up in the subcompartments outside of the containment for postulated breaks are the same as those used for subcompartments inside the containment. Also, the allowance for structural design margins (pressure) should be the same. If different methods are used, justify that your method provides adequate design margins and identify the margins that are available. When you submit the results of your evaluation, identify the computer codes used, the assumptions used for mass and energy release rates, and sufficient design data so that we may perform independent calculations.

RESPONSE:

The requested information is provided in Appendix 3.6A.

SSES-FSAR Rev. 46, 06/93 010.3-1 QUESTION 010.3 The peak pressures and temperatures resulting from the postulated break of a high energy pipe located in compartments or buildings is dependent on the mass and energy flows during the time of the break. You have not provided the information necessary to determine what terminates the blowdown or to determine the length of time blowdown exists. For each pipe break or leakage crack analyzed, provide the total blowdown time and the mechanism used to terminate or limit the blowdown time of flow so that the environmental effects will not affect safe shutdown of the facility.

RESPONSE:

For those pipe breaks analyzed, termination of blowdown was not a controlling factor in the analysis since the temperature and pressure peaked within the first few seconds after the line break. Short term blowdown in these cases does not result in higher temperatures and pressure. Termination of the blowdown for breaks outside containment is accomplished by an automatic isolation signal from the Leak Detection System described in Subsection 7.6.1a.4.

SSES-FSAR Rev. 51, 02/97 010.4-1 QUESTION 010.4 The design criteria for the main steam isolation valve leakage control system (MSIVLCS) does not contain provisions to prevent the operation of the MSIVLCS when the inboard MSIV fails to close. We will require that an additional interlock be provided on the main steam isolation valve leakage control system so that the operation of an inboard leakage control system is prevented should an inboard main steam line isolation valve fail to be in its fully closed position.

RESPONSE:

The system design basis considers that there will be appreciable hold up time following the design basis LOCA before fission products from the core are transported down the main steam lines. The leakage control system is designed so that if the inboard system is actuated with one inboard MSIV failed open, the vent line from that steam line will automatically reclose by the time fission products, assuming plug flow, move down the steam line about 1/2 the total pipe run from the reactor vessel to the failed open inboard MSIV. Further, the operator is afforded with information from the control room regarding the status of the MSIV's through the valve position switches. Operating procedure inhibits the operator in activating the system in the event the valve position indicator shows failed open. Therefore, it is concluded that the additional interlock is not warranted.

NOTE: MSIV-LCS information maintained here for historical purposes. The MSIV-LCS has been deleted. The function is now performed by the Isolated Condenser Treatment Method (Section 6.7).

SSES-FSAR Rev. 46, 06/93 010.5-1 QUESTION 010.5 You state that the Unit 1 facilities reactor building crane is a single failure proof crane and is designed to handle the spent fuel cask, and that the Unit 2 crane is not single failure-proof and is designed to handle all normal plant operation loads except the spent fuel cask. Provide the following information for these fuel handling systems:

(1) Describe the normal plant operation loads that the Unit 2 reactor building crane is capable of carrying in the fuel building area.

(2) Describe the means used to prevent the Unit 2 reactor building crane from handling the spent fuel cask when stored in the spent fuel shipping cask storage pool. (3) Describe the mechanical stops and/or electrical interlocks that would restrict the path of the 125-tone crane to those areas identified on Figure 9.1-16A and 9.1-16B.

(4) State whether the Unit 1 reactor building crane has been designed to meet the guidelines of Branch Technical Position ASB 9-1, "Overhead Handling Systems for Nuclear Power Plants." RESPONSE:

1) Please see revised Subsection 9.1.5 for this information.

2) Please see revised Subsection 9.1.5 for this information.

3) Please see revised Subsection 9.1.5.3 for this information.

4) See response to Question 010.25.

SSES-FSAR Rev. 51, 02/97 010.6-1 QUESTION 010.6 A single failure of an inboard MSLIV would allow a continuous blowdown of the containment atmosphere to the reactor building standby gas treatment system for a specified period of time when the MSIVLCS is initially actuated. This violates our containment isolation criteria and the consequences of the blowdown are unacceptable. It is our position that an interlock be provided so that the leakage control system actuation valves can be opened only if the associated inboard MSLIV is in a fully closed position. Revise the FSAR to indicate conformance to our position.

RESPONSE:

The system design basis from the outset has been that there will be appreciable hold up time following the design basis LOCA before fission products from the core are transported down the main steam lines. The leakage control system is designed so that if the inboard system is actuated with one inboard MSIV failed open, the vent line from that steam line will automatically reclose by the time fission products, assuming plug flow, move down the steam line about 1/2 the total pipe run from the reactor vessel to the failed open inboard MSIV. Besides, the operator is afforded with information from the control room regarding the status of the MSIV's through the valve position switches. Operating procedures inhibit the operator in activating the system in the event the valve position indicators show failed open. We conclude that the additional interlock is not warranted.

NOTE: MSIV-LCS information maintained here for historical purposes. The MSIV-LCS has been deleted. The function is now performed by the Isolated Condenser Treatment Method (Section 6.7).

SSES-FSAR Rev. 51, 02/97 010.7-1 QUESTION 010.7 The design criteria for the main steam isolation valve leakage control system indicates that you propose to allow a main steam isolation valve (MSIV) leakage rate up to 100 SCFH for each MSIV in each steamline. It is our position that the design basis leak rate of 100 SCFH is not an acceptable MSIV leakage rate for normal operation. Therefore, we will still impose a technical specification limit of 11.5 SCFH for the MSIV leak rate and a leak rate verification testing frequency consistent with the plant Technical Specifications used for other operating BWR's. Revise the FSAR to indicate that the MSIV leak rate for normal operation will be limited to 11.5 SCFH.

RESPONSE:

It is stated in Section 6.7.1.3 of the FSAR that the main steam isolation valve leakage control system (MSIV-LCS) is designed to process MSIV leakage rates up to 100 SCFH for each MSIV in each line. This is a design basis for the MSIV-LSC and is not the design basis leakage rate for the MSIV's. The Standard Technical Specification in Chapter 16 of the FSAR specifies the MSIV leakage rate at 11.5 scf per hour.

NOTE: MSIV-LCS information maintained here for historical purposes. The MSIV-LCS has been deleted. The function is now performed by the Isolated Condenser Treatment Method (Section 6.7).

SSES-FSAR Rev. 46, 06/93 010.8-1 QUESTION 010.8 Confirm that a Keff of less than 0.98 will be maintained with fuel of the highest anticipated reactivity in place in the new fuel storage racks and assuming optimum moderation.

RESPONSE:

See revised FSAR Subsections 9.1.1.1.1.2, 9.1.1.2 and 9.1.1.3.1.

SSES-FSAR Rev. 46, 06/93 010.9-1 QUESTION 010.9 The information contained in the Susquehanna FSAR is not of sufficient detail to support a conclusion that the liner plate for the spent fuel pool is designed to seismic category I. Therefore, we require, that you demonstrate compliance with Regulatory Guides 1.13 and 1.29 by showing that a failure of the liner plate as a result of an SSE will not affect any of the following: significant release of radioactive materials due to mechanical damage to the spent fuel; significant loss of water from the pool which could uncover the fuel and lead to release of radioactivity due to heat-up; loss of ability to cool the fuel due to flow blockage caused by a portion or one complete section of the liner plate falling on top of the fuel racks; damage to safety related equipment as a result of pool leakage; or uncontrolled release of significant quantities of radioactive fluids to the environs.

RESPONSE:

See revised Subsections 9.1.2.1 and 9.1.2.2.

SSES-FSAR Rev. 46, 06/93 010.10-1 QUESTION 010.10 Confirm that all portions of the structure (reactor building) which serve as a low leakage barrier to provide atmospheric isolation of the spent fuel storage pool and associated fuel handling area are designed to seismic Category I criteria.

RESPONSE:

See revised FSAR Subsection 9.1.2.2.

SSES-FSAR Question Rev. 47 FSAR Rev. 64 010.11-1 QUESTION 010.11

The spent fuel pool cooling syst em is a non-seismic system.

This does not meet the guidelines set forth in Regulatory Guide 1.

13 and 1.29. Analyze t he design of the spent fuel pool cooling system to show that the pumps and piping are supported so that they are capable of withstanding an SSE, or provide the results of an analysis to show that for the complete loss of fuel pool cooling that would result in pool boiling, a release of significant quantities of radioactivity to the environment will not result.

RESPONSE:

A complete analysis showing the amount of radioactive release following a complete loss of fuel pool cooling is provided in Appendix 9-A. As shown in Table 9A-1 the thyroid dose consequences of the boiling pool are well below the guideline values of 10CFR50.67 and the 0.5 REM TEDE thyroid guideline.

Subsection 9.1.2.2 provides the logic which shows that the spent fuel pool will not drain following an SSE.

SSES-FSAR Rev. 46, 06/93 010.12-1 QUESTION 010.12 Confirm that a spent fuel pool water temperature of 125F is maintained when the fuel pool cooling system is used to cool the emergency heat load.

RESPONSE:

A spent fuel pool water temperature of 125F is maintained when the fuel pool cooling and cleanup system (FPCCS) is used in conjunction with the RHR cooling system to cool the emergency heat load. Refer to revised section 9.1.3.1 for FPCCS design basis.

SSES-FSAR Rev. 51, 02/97 021.1-1 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:

SSES-FSAR Rev. 51, 02/97 021.2-1 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.

SSES-FSAR

Rev. 51, 02/97 021.19-1 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.

SSES-FSAR

Rev. 51, 02/97 021.26-1 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.

SSES-FSAR

Rev. 51, 02/97 021.39-1 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.

SSES-FSAR

FSAR Rev. 68 021.67-1 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. The heat-up time and the initiation time for the hydrogen recombiners are discussed in Subsection 6.2.5.2. Note the hydrogen recombiners are no longer

credited in the accident analysis and do not have a safety related function. The

equipment is still maintained safety related.

SSES-FSAR Rev. 51, 02/97 021.73-1 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.

SSES-FSAR Rev. 51, 02/97 021.75-1 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.

AutoCAD: Figure Fsar 032_11_1.dwgFSAR REV.65FIGURE 032.11-1, Rev 47ADS VALVE CONTROLSUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 & 2FINAL SAFETY ANALYSIS REPORT

SSES-FSAR Question Rev. 47 FSAR Rev. 64 032.22-1 QUESTION 032.22 The standby liquid control system (SLCS) is designated as a special safety system in the SSES design. To assure that availability of the SLCS, you have provided two sets of the components required to actuate the system in parallel redundancy. However, our review indicates that you have not provided redundant heating systems and the heating equipment supplied is not powered from an emergency bus under normal conditions. The staff has concluded, therefore, that the statement in FSAR Section 9.3.5.3 that a "single failure will not prevent system operation" is not true. Provide a modified design of the SLCS which satisfies the single failure criterion or justify the present design.

RESPONSE: Regulatory Guide 1.70 identifies the SLCS as a safe shutdown system having a safety-related classification. Safety-related systems provide the actions necessary to assure safe shutdown of the reactor, to protect the integrity of radioactive material barriers, and/or to prevent the release of radioactive material in excess of allowable dose limits. However, the system fails to meet all the requirements of a safety-related system in that SLCS is not designed for the single active component failure criteria. A function of the system is to inject boron to the suppression pool via the SLC System to maintain basic pH in the suppression pool in order to minimize re-evolution of iodine from the suppression pool in the event of a Loss of Coolant Accident (LOCA). This function adds additional importance to the system's performance.

The NRC has provided review guidelines for the SLC system that do not meet single failure criteria or that are not of the expected quality (safety related). To demonstrate that the SLC System is able to perform its AST (10 CFR 50.67) function (injection of sodium pentaborate into the suppression pool), the System should satisfy, as a minimum, the recommended guidelines listed below (NRC review guidelines). Meet ing these criteria, demonstrates reasonable assurance of the quality of the SBLC System. These guidelines are as follows:

a) The SLC System should be provided with standby AC power supplemented by the emergency diesel generators.

b) The SLC System should be seismically qualified in accordance with Regulatory Guide 1.29 and Appendix A to 10 CFR Part 100.

c) The SCL System should be incorporated into the plant's ASME Code ISI and IST Programs based upon the plant's code or record (10 CFR 50.55a).

d) The SCL System should be incorporated into the plant's Maintenance Rule program consistent with 10 CFR 50.65 e) The SLC System should meet 10 CFR 50.49 and Appendix A (GDC 4) to 10 CFR 50.

An extensive validation of NRC guidelines for the SBLC system of components that do not meet single failure criteria or that are not of the expected quality (safety related) provides reasonable assurance of the System's ability to support its original and the pH controlling functions.

SSES-FSAR Question Rev. 47 FSAR Rev. 64 032.22-2 Subsection 9.3.5.3 discusses the consequences of a loss of the tank heaters or suction piping heat tracing. Note that, as stated in Subsection 7.4.1.2.2, the tank heaters receive power from the standby a-c power supply bus C (Division I). The design has been changed to provide Division I standby a-c power to the heat tracing circuit as well.

AutoCAD: Figure Fsar 032_57_1.dwgFSAR REV.65FIGURE 032.57-1, Rev 47MSIV PERCENT CLOSUREVS.TIME EXTRAPOLATIONSUSQUEHANNA STEAM ELECTRIC STATIONUNITS 1 & 2FINAL SAFETY ANALYSIS REPORT

SSES-FSARQuestionRev.48FSARRev.65032.74-1QUESTION032.74ForthePCRVICS,alargenumberofinconsistencies,errors,omissions,andconflictswerenotedbetweenthedescriptions(7.3.1.1a.2),theanalyses(7.3.2a.2),thefunctionalcontroldiagram(Figure7.3-8)andtheelementarydiagrams(791E401AE,791E414AE,and791E425AE).Someexamplesfollow:1)TheFSAR(7.3.1.1a.2.4.1.1.1)indicatesfourlevelswitcheswithtwosetsofcontactseach-onesetofcontactsforlowlevelandonesetforlowlow(lower)level.Also,asinglepairofreactorvesselpressuretapsforeachpairofswitcheswasindicated.Figures5.1-3bandFigure7.3-8andelementarydiagrams791E401AEand791E414AEshowtwosetsoffoureachlevelswitches-oneforlowlevelandoneforlowlowlevel.Figure5.1-3balsoshowsthelowlevelandlowlowlevelswitchesconnectedtodifferencepressuretaps.2)TheFSAR(7.3.1.1a.2.4.1.1.1)indicatesthatthelowlowwaterlevelsignalisolatestheMSIVs,thesteamlinedrainvalves,thesamplelines,and"allotherNSSSisolationvalves."FurtherreviewoftheFSARtext,figures,andelementarydiagramsshowslowlowwaterlevelonlyisolatestheMSIVs,steamlinedrainvalves,andthesamplelines.No"otherNSSSisolationvalves"couldbefoundthatwereactuatedbythelowlowwaterlevelsignal.3)TheFSARindicatesthePCRVICSinstrumentationandcontrolsubsystemsinclude:(10)mainsteamline-leakdetection;(12)reactorwatercleanupsystem-highflow,(14)reactorcoreisolationcoolingsystem-highflow,and(15)highpressurecoolantinjectionsystem-highflow.Theremainingtextdoesnotdiscusstheseitemsnorweretheyfoundintheelementarydiagramsorfigures.4)TheFSAR(7.3.1.1a.2.4.1.9)indicatesthatRWCUsystemhighdifferentialflowissensedwith"twodifferentialflowsensingcircuits"andtheanalysessectionindicatesthePCRVICScompliesfullywiththesinglefailurecriteria.TheRWCUP&IDandtheelementarydiagramsshowonlyonehighdifferentialflowinstrumentconsistingofthreeflowtransmittersdrivingasinglesummerwhich,inturn,drivestwoalarmunits(oneforeachofthetwotripchannels).Thisarrangementdoesnotmeetthesinglefailurecriteria.5)Elementarydiagram791E401AEshowsadevice(dPISG33-NO44A)labeled"HighDiffFlow"inadditiontothedevicein4)above.N044AappearsasadifferentialpressureswitchintheRWCUP&ID.Nootherreferencetothisdevicecouldbefoundinthetextorelementarydiagrams.6)Thetextstates"RWCUsystemhighdifferentialflowtripisbypassedautomaticallyduringRWCUsystemstartup."Noinformationonthisbypasscouldbefoundinthetextorelementarydiagrams(791E401AEor791E423AE)orinthevariousanalysesinSection7.3.2a.2.

SSES-FSARQuestionRev.48FSARRev.65032.74-27)Thetextindicates"maincondenserlowvacuumtripcanbebypassedmanuallywhentheturbinestopvalveislessthan90%open."Elementarydiagram791E401AEandtheresponsetoQ032.33showsthat"reactorlowpressure"isalsorequiredtoallowthisbypass.Nootherinformationonthis"reactorlowpressure"permissive,includingthesetpoint,couldbefoundintheFSAR.8)TheFSAR(7.3.1.1a.2.5and7.3.1.1a.2.11)mentionsa"highdifferentialpressure"signalusedforRWCUisolation.Nootherinformationcouldbefoundonthissignaleitherinthetextortheelementarydiagrams.9)TheFSAR(7.3.1.1a.2.11)mentions"hightemperaturedownstreamofthenon-regenerativeheatexchanger"asaRWCUisolationsignal.Elementarydiagram791E401AEalsoshowsthissignal,butonlyshowsasingleinstrument,whichdoesnotmeetthesinglefailurecriteria.Thisisolationsignalisnotdiscussed,described,orjustifiedinthetextortheanalyses.10)Elementarydiagram791E401AEalsoshowsasingleSBLCsystemisolationsignalthatdoesnotmeetsinglefailurecriteria.Thissignalisalsonotdiscussedinthetextortheanalyses.11)Elementarydiagram791E401AEshowsanRHRisolationfor"ExcessFlow"and"HighReactorPressure".Noinformationcouldbefoundonthesesignalsinthetextoranalyses.12)ThetextindicatesthatRWCUandRHRsystemshighareaanddifferentialtemperaturesubsystemshave"noautomaticbypasses."Elementarydiagram791E401AEshowsamanualbypassswitchforthissybsystem.ThetextalsosaysthatthemainsteamlinelowpressureandthecondenserlowvacuumbypassesaretheonlybypassesinthePCRVICS.13)ThetextindicatesthatthemainsteamlinehighradiationsystemhasbypassesontheindividualinstrumentsthatarenotdescribedintheFSARorincludedintheanalyses(7.3.2a.2).Amendtheappropriatedocument(s)tofullyandaccuratelydescribethePCRVICSinstrumentationandcontrolsystemsactuallyinstalledatSusquehanna.AmendthePCRVICSanalysespresentedinSection7.3.2a.2toagreewiththesystemsdiscussedinthetextandshowninthefiguresandelementarydiagrams.Forthebypasses,fullydescribethejustifyallmanualorautomaticbypassesassociatedwithanyPCRVICSsubsystemandincludeallbypassesinthevariousSection7.3.2a.2analyses.Includeadescriptionofhowallbypassesareannunciated.Also,reviewthecompletePCRVICSdescriptionsandanalysesgivenintheFSARandthefiguresandelementarydiagrams.VerifythatthesedocumentsaccuratelydescribethesystemsactuallyinstalledatSusquehanna.

SSES-FSARQuestionRev.48FSARRev.65032.74-3RESPONSE:1)Subsection7.3.1.1a.2.4.1.1.1hasbeenamendedtoindicateonesetoffourlevelswitchesisforlowlevelandasecondsetisforlowlowlevel.Thereisonecommonandtwovariablelegpressuretapsforeachpairoftwowaterlevels.2)Subsection7.3.1.1a.2.4.1.1.1hasbeenmodifiedtoread;"thesecond(andlower)......isolationvalvesandotherselectedisolationvalves.IsolationvalvesandtheirinitiatingsignalsareshowninTable6.2-12."TheNuclearBoilerFCDandNS 4elementarydiagramwillbemodifiedtoremovedrywellpressureasaninitiatingsignalfromtheRHRisolationvalves,excepttheRadwastedischargevalvesandtheheatexchangervalves.DrywellpressureasaninitiatingsignalwillalsoberemovedasaninputtotheRWCUvalves.Inaddition,level3(lowlevel)isolationwillbechangedtolevel2(lowlowlevel)forallvalvesexcepttheRHRandTIP.ThesechangestothePCRVICSinitiationsignalsandsetpointshavenoimpactonsafety.Theyhavebeenimplementedasaplantavailabilityfeaturei.e.toreduceinadvertentcontainmentisolationduringplanttransientevents.ThePCRVICSFCDandelementarydiagramhasbeenupdatedtoreflectthesechanges.3)MainsteamlineleakdetectionisdiscussedinSubsection7.3.1.1a.2.4.1.12.RCIChighflowandHPCIhighflowarediscussedintheresponsetoQuestion032.55.4)Thesinglefailurecriterionappliesatthesystem(RPS)orfunction(ECCS)levelandnotatthesignalinputorchannellevel.TheRWCUisolationvalveswillreceiveasystemisolationsignalfromthespacetemperaturetripchannelsandhighflowsignaldescribedin(5)belowifabreachoccursintheRWCUsystemRCPBandtheflowsummerfailed.SinglefailureofthesummerwillnotprecludeRWCUsystemisolation.5)G33-NO44AandBprovideanRWCUsystemisolationsignalonhighflowinthesuctionline.AreviseddiscussioniscontainedinSubsection7.3.1.1a.2.4.1.9.6)Subsection7.3.1.1a.2.4.1.9.6hasbeenmodifiedtostatedthattheRWCUsystemhighdifferentialflowtripisbypassedduringRWCUsystemstartupbyatimedelay.ThetimedelaywillnotaffectRWCUSystemRCPBisolation.WhentheRWCUsystemisinitiatedahighdifferentialflowwillexistbetweentheinletandoutletflowsandinitiatesystemisolationandpreventRWCUoperations.Thetimedelaybypassestheflowsignaluntilthesystemloopflowisestablished.7)SeerevisedSubsection7.3.1.1a.2.4.13.6 SSES-FSARQuestionRev.48FSARRev.65032.74-48)SeerevisedSubsections7.3.1.1a.2.5and7.3.1.1a.2.11whichusetheterm"highflow"whichisnowdescribedperpart(5)above.9)Thesubjectsignalisasystemtripsignal,notacontainmentisolationsignal.Seetheresponsetopart4andseeSubsection7.7.1.8.2.2(1)and7.7.2.8.1forcoverageofthissignal.10)ThesignalisrequiredforSLCSoperationandisnotacontainmentisolationsignal.TheRWCUSystemwillbemanuallyshutdown,ifstandbyliquidinjectionisrequired,topreventboronlosstotheRWCUsystem.Seeresponsetopart4andseerevisedSubsection7.4.1.2.5.1tocovertheneedformanualisolationofRWCU.11)ExcessflowisdiscussedundertheLeakDetectionSystem,inSubsection7.6.1a.4.3.5.3.HighreactorpressureisdiscussedunderHighPressure/LowPressureSystemInterlocks,inSubsection7.6.1a.3.3.1.12)ThetextinSubsections7.3.1.1a.2.4.1.10.6and7.3.1.1a.2.4.1.11.6isreferringtooperatingbypasses,whicharethesubjectofdiscussioninSubsection7.3.2a.2.2.3.1.1.2.TheManualBypassSwitchesareshownontheLeakDetectionElementaryDiagramforRWCUandRHR.Theseswitchesactuatethesystemintestannunciator.TestandmaintenancebypassesarealsodiscussedinRegulatoryGuide1.47conformance(seeSubsection7.3.2a.2.2.1.5).13)Subsection7.3.1.1a.2.4.1.2.5discussesbypassesandstatesthattherearenooperationalbypassesprovidedwiththeMainSteamlineHighRadiationMonitoringSubsystem.Theindividuallogradiationmonitorsmaybebypassedformaintenanceorcalibrationbytheuseoftestswitchesoneachmonitor.Bypassingonelogradiationmonitorwillnotcauseanisolationbutwillcauseasingletripsystemtriptooccur.

SSES-FSARQuestionRev.53FSARRev.65040.41-1QUESTION040.41Discusstheprecautionarymeasuresthatwillbetakentoassurethequalityandreliabilityofthefueloilsupplyforemergencydieselgeneratoroperation.Includethetypeoffueloil,impurityandqualitylimitationsaswellasdieselindexnumberofitsequivalent,cloudpoint,entrainedmoisture,sulfur,particulatesandotherdeleteriousinsolublesubstances;procedurefortesting newlydeliveredfuel,periodicsamplingandtestingofon-sitefueloil(includingintervalbetweentests),intervaloftimebetweenperiodicremovalofcondensatefromfueltanksandperiodicsysteminspection.Inyourdiscussionincludereferencetospecificindustry(orotherstandards whichwillbefollowedtoassureareliablefueloilsupplytotheemergencygenerators.RESPONSE:DiscussionincludedinSection9.5.4.4oftheFSAR.