ML20148M616

Forwards Response to Question 032.1 Re Staff Positions & Questions Transmitted to All Appl w/RESAR-3 Plants on Instru & Control Sys

ML20148M616
Person / Time
Site:	Comanche Peak
Issue date:	11/16/1978
From:	Feist C TEXAS UTILITIES SERVICES, INC.
To:	Naventi R Office of Nuclear Reactor Regulation
References
NUDOCS 7811220083
Download: ML20148M616 (49)

Text

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TEXAS UTILITIES SERVICES INC.

2tXJi DHYAN TO% EH

• DAI.1,AR TEX All 7M201 TXX-2912 November 16, 1978 Mr. Ron Naventi Licensing Project Manager Light Water Reactors Branch No. 4 Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.

20555 COMANCHE PEAK STEAM ELECTRIC STATION TRANSMITTAL OF MATERIAL REQUESTED BY QUESTION 032.1 FILE NO. 10010

Dear Mr. Naventi:

Please find enclosed three (3) copies of our response to

,Y question 032.1 concerning staff positions and questions transmitted to all applicants with RESAR-3 plants.

Sincerely,

0. l(. h C. K. Feist CKF:skf Enclosure cc:

N. S. Reynolds, Esq. w/o enclosures S. C. Relyea, Es4. w/o enclosures H. C. Schmidt w/o enclosures H. R. Rock w/ enclosures G. L. Hohmann w/ enclosures J. T. Merritt w/ enclosures J. C. Kuykendall w/ enclosures 781122008%

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'CPSES/FSAR 03?.0 IllSTRUMEtiTAT10tl At1D C0ilTROL SYSTEMS BRAtlCil l

QO32.1 Provide a listing of the sections of the FSAR which include the responses to the staff positions and questions transmitted to all applicants with RESAR-3 plants.

This transmittal was dated liovember 17, 1977.

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R032.1 The set of RESAR-3 questions and responses follows.

flote I

that all references to Sections and Questions in the Responses are to the CPSES/FSAR.

i RESAR-3 Section 3.9.1.2 of RESAR-3 states that dynamic testing (Q31.1) procedures concerning Westinghouse supplied safety-related mechanical equipment will be provided in the applicant's FSAR.

It is our position that as a minimum you commit to conduct a seismic qualification program to conform to the criteria as contained in Attachment A.

State your intent

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to employ the criteria as contained in Attachment A for all Westinghouse Category I mechanical equipaent in order to confirm the functional operability of such equipment during and after a seismic event up to and including the SSE.

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Response

See Subsection 3.9ti.2.2 4

RESAR-3 Section 3.9.2.4.1 of RESAR-3 states that the pump motor and l

(Q31.2) vital auxiliary electrical equipaent will be qualified by meeting the requirements of IEEE Standard 344-1971.

Since the standard has undergone a major revision, state your intent to meet the requirements of the 1975 version of IEEE Standard 344.

IEEE Standard 344-1975 includes requirements which are applicable to all plants with C.P. applications l

docketed af ter October 1972.

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1 032-1 I

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CPSES/FSAR

Response

See Question 112.9 and Subsection 3.9N.3.2 I

RESAR-3 The Seismic qualification criteria for electrical equipment j

(Q31.3) as stated in Section 3.10 of the proposed Amendment 6 to RESAR-3 is not completely acceptable because it is only l

applicable to certain specific conditions when single I

, frequency input to an individual axis is justifiable.

A 1

broader criterion to account for overall considerations l

should be provided.

The major concern is the possible 1

directional coupling and the concurrent multi-mode 1

I response.

An acceptable response is to conduct a seismic j

qualification program as reconinended by the 1975 version of IEEE-344 standard.

State your intent to use this i

reconnended criteria.

I 1

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Response

See Section 3.10N l

RESAR-3 The lists of safety-related equipment-and components pro-(Q31.4) vided in Section 3.11.1 of RESAR-3 are not complete.

Identify all individual components and complete the lists.

Response

See Section 3.11N RESAR-3 Section 3.11.2 of RESAR-3 does not give a complete and (Q31.5) acceptable description of the qualification tests and analyseo for each type of safety-related equipment and component.

Provide this information for each item.

Response

See Section 3.11N RESAR-3 RESAR-3 Section 7.1.2.5.

Describe how your design complies (Q31.6) with IEEE Standard 323-1971, or IEEE Standard 323-1974, for all applications for which the construction pennit safety evaluation report was issued July 1,1974 or later.

1 032-la

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CPSES/FSAR Identify and justify all exceptions.

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Response

See Section 3.11N RESAR-3 in accordance with the implementation da'ces (noted in

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(Q31.7) parentheses) and as they apply to your application, describe the extent to which the recommendations of the following regulatory guides will be met.

Identify and justify any exception.

Regulatory Guide 1.22 (Safety Guide 22), " Periodic Testing f

a of Protection System Actuation Functions" (Guide dated i

2/17/72)

Regulatory Guide 1.29, " Seismic Design Classification;"

(Revision 1 dated August 1973)

Regulatory Guide 1.30 (Saftey Guide 30), " quality Assurance Requirements for the Installation, Inspection, and Testing of Instrtmentation and Electric Equipment;" (Guide dated August 11,1972)

Regulatory Guide 1.40, " Qualification Tests of Continuous-Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants;" (Guide dated 3/16/73)

Regulatory Guide 1.47, " Bypassed and Inoperable State Indication for Nuclear Power Plant Safety Systems;" (Guide dated May 1973)

Regulatory Guide 1.53, " Application of the Single-Failure Criterion to Nuclear Power Plant Protection System;" (Guide dated June 1973) 032-1b

t CPSES/FSAR Regulatory Guide 1.62, " Manual Initiat. ion of Protection Action;" (Guide dated October 1973)

Regulatory Guide 1.63, " Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants;" (Guide dated October 1973)

Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors;" (Guide l

I dated November 1973)

Regulatory Guide 1.73, " Qualification Tests of Electric Valve Operators Installed Inside the containnent of Nuclear Power Plants;" (Guide dated January 1974)

Regulatory Guide.1.75, " Physical Independence of Electric Sy s t ems."

The physical indentification of safety-related equipment should also be addressed in this section; (Guide dated February 1974)

Regulatory Guide 1.80, "Preoperational Testing of Instrmuent Air System;" (Guide dated June 1974) and Regulatory Guide 1.39, " Qualification of Class IE Equipnent for Nuclear Pcwer Plants."

(Applicable to all plants with an SER issued af ter July 1,1974).

032-Ic l

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CPSES/FSAR

Response

Regulatory Guide FSAR Section 1.22 Appendix 1A(N) 1.29 Appendix 1A(N) & 1A(B) 1.30 Appendix 1A(N) & 1A(B) 1.40 Appendix 1A(B) 1.47 Appendix 1A(B) 1.53 Appendix 1A(N) 1.62 Appendix 1A(N) 1.63 Appendix 1A(B) 1.68 Appendix 1A(B) 1.73 Appendix 1A(B) & 1A(N) 1.75 Appendix 1A(N) & 1A(B) 1.80 Appendix 1A(B) 1.89 Appendix 1A(N) & 1A(B)

RESAR-3 Provide a discussion and the results of an analysis showing j

(Q31.8.1) how your design of the test and calibration features of the safety systems meets the requirements of Section 4.10 of IEEE Standard 279-1971.

Response

See Subsections 7.1.2.5, 7.1.2.11 and Figure 7.3-1, 7.3-3.

RESAR-3 Based on figure 7.2-1, Sheet 7 of 17, of RESAR-3, we have (Q31.8.2) concluded that the proposed design for the steamline differential pressure circuits does not conform to the requirenents of IEEE Standard 279-1971.

Specifically, during operation with a loop isolated, the logic for the operable steamlines is effectively changed to 2-out-of-2 which does not meet the single failure criterion.

Our position is that in order to comply with IEEE Standard 279-1971, the desing should incorporate positive means of 032-1d

CPSES/FSAR J

assuring that these circuits continue to meet the single failure criterion during operation with a coolant loop isolated. Discuss your intent' to comply wiui this position and describe the necessary design changes, or justify any exceptions by discussing your reasons for concluding that such exceptions are in accordance with the requiren.3nts of IEEE Standard 279-1971.

In addition as conmitted on Fage 1

7.2-30 of RESAR-3, provide the results of an analysis that will detennine whether automatic tripping of the steamline i

differential pressure bistables is required for N-1 loops operating.

Response

Steanline break protection sensors and logic are being changed. The revised design comaitment will be provided in the last quarter of 1978.

RESAR-3 RESAR-3 Section 7.2.1.1.2(1)(d) and Figure 7.2-1 Sheet 3 (Q31 9) address a power range high neutron flux rate " Positive" trip. This trip is used as protection against a rod ejection accident.

The referenced Westinghouse Topical Report WCAP-7380-L (pages 2-8 and 3-12) provides a diagram and a description for the " Negative" flux rate trip but does not provide for the " Positive" flux rate trip.

Provide a description and diagram covering " Positive" flux rate trip.

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Response

WCAP-7380-L is no longer referenced.

It is replaced with WCAP-8255 as listed in the references for FSAR Subsection 7.2.

WCAP-8255 discussed both the positive and neg6tive rate trips and provides diagra'as for both.

RESAR-3 The reactor trip system contains logic circuits that can i

(Q31.10) initiate trips for the purpose of anticipating the approach to a limiting condition for operation.

Sepcifically, these 032-le i

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~, - - - _ _ _ _ _ _ _ _ _ _. - _ - _ _ - - _ -

CPSES/FSAR reactor trips are:

1 (1)

Generation of a reactor trip by tripping the main coolant punp breakers, (2)

Generation of a reactor trip by tripping the turbine, (3)

Generation of reactor trip by underfrequency conditions on reacotr coolant paup bus, and (4)

Generation of reactor trip by undervoltage conditions on reactor coolant punp bus.

Our position requires that all inputs to the reactor trip system be designed to meet IEEE Standard 279-1971, with an exception for anticipatory trips (trips not required for safety actions in the accident analysis - Chapter 13).

The exception is that sensors for anticipatory trips are not required to be located in a qualified seismic Category I structure.

Discuss your intent to canply with this position or justify any exceptions you may have in this regard.

Your response should include a discussion of the testability of.these circuits while the reactor is at power.

Response

(1)

The design is changed. The reactor trips by main coolant punp breaker opening is one condition of the undervoltage trip.

See Subsection 7.2.1.1.2(4b).

(2)

See Subsection 7.2.1.1.2(6)

(3)

See Subsection 7.2.1.1.2(4c)

(4)

See Subsection 7.2.1.1.2(4b) 032-1f

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CPSES/FSAR For a discussion of testability of these circuits while the reactor is at power, refer to Section 7.1.2.5.

1 RESAR-3 Testing of the reactor trip system and the engineered (Q31.11) safety feature actuation system to verify that the system" response times are equal to or less than the values assuned in the accident analysis is discussed on Pages 7.1-19, 7.2-24, and 7.3-13 or RESAR-3.

In addition to the proposed response time testing during preoperational start-up testing and following the replacanent of a canponent that affects response time, our position requires that these systens be designed to permit periodic verification that i

the response times are within the values assuned in the accident analysis.

Discuss your intent to canply with this position or justify any exceptions.

It is stated in RESAR-3 on Pages 7.3-26 that the response time specified in Paragraph 4.1 of IEEE Standard 338-1971 is not checked periodically as is the set point accuracy.

Provide justification for.the exception to this requirenent.

Response

See Subsection 7.1.2.11 RESAR-3 With regard to the motor operated accumulator isolation (Q31.12) values, we require.that the proposed design include the following features in order to conform to the requirenents of IEEE Standard 279-1971:

(1)

Autaaatic opening of the accumulator valves when either (a) the primary coolant system pressure exceeds a preselected value (to be specified in the Technical Specifications) or (b) a safety injection signal has been initiated.

Both signals shall be provided to the values.

032-1g t

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CPSES/FSAR (2)

Visual indication in the control room of the open or closed status of the valve, actuated by sensors on the valve.

(3)

An audible alarm, independent of Item (2), that is actuated by a sensor on the valve when the valve is not in the fully open position.

(4)

Utilization of a safety injection signal to autanatically renove (override) any bypass feature that may be provided to allow an isolation valve to be closed for short periods of time when the reactor coolant system is at pressure (in accordance with the provisions of the proposed Technical Specifications).

Discuss your intent to canply with these requirenents or justify any exceptions to these requirements.

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Response

See Subsection 7.6.4 RESAR-3 Based on the information provided in Secton 7.3 of RESAR-3, (031.13) we conclude that the proposed design for manual initiation of steamline isolation does not confona viith the requirenents of Section 4.17 of.IEEE Standard 279-1971.

In addition, there is not sufficient infonuation on the design provision for manual initiation of containnent isolation and containnent depressurization to determine whether these functions are desinged in accordance with.Section 4.17 of IEEE Standard 279-1971.

Our position is that a design which meets the following is an acceptable means of meeting the requirenents of Section 4.17 of IEEE Standard 279-1971:

(1)

Means should be provided for manual initiation of each protective action (e.g., reactor trip, containnunt isolation) at the system level, 032-lh 1

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CPSES/FSAR regardless of whether or not means are also provided to initiate the protective action at the component or channel level (e.g., individual control rod, individual isolation valve).

(2)

Manual initiation of a protective action at the system level should perfonn all actions performed by autoaatic initiation such as starting auxiliary or supporting systen, sending signals to appropriate valves to assure their correct position, and providing the required action-sequencing functions and interlocks.

(3)

The switches for manual initiation of protective actions at the system level should be located in the control room and be easily accessible to the operator so that action can be taken in an expeditious manner.

(4)

The anount of equignent common 'to both manual and autonatic initiation should be kept to a minimum.

It is preferable to limit such common equipnent to the final actuation devices and the actuated equipnent.

However, action-sequencing functions and interlocks

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(of Position 2) associated with the final actuation devices and actual equipment may be common providing individual manual initiation at the component or channel level is provided in the control roan. No single failure within the manual, autonatic, or conmon portions of the protection systen should

. prevent initation of protective action by manual or autonatic means.

(5)

Manual i'nitiation of protective actions should depend on the operation of a minimum of equipnent consistent 032-li l

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CPSES/FSAR with 1, 2, 3, and 4 above.

(6)

Manual initiation of protective action at the system level should be so designed that once initiated, it will go to completion as required in Section 4.16 of IEEE Standard 279-1971.

Discuss your intent to comply with this position or justify any exceptions by discussing your reasons for concluding that such exceptions are in accordance with the requirenents of IEEE Standard 279-1971.

Response

See Subsection 7.3.2.2.7 RESAR-3 General Design Criterion 37 requires, in part, that the (Q31.14) emergency core cooling system by designed to pennit testing the operability of the system as a whole.

On Page 7.3-26 of RESAR-3, it is stated that the safety injection and residual heat removal pumps are made inoperable during the system tests.

Our position is that in order to comply with the requirements of Criterion 37, these pumps must be included in the system test.

Discuss your intent to comply with this position or justify any exception.

Response

See Subsection 6.3.4 and Appendix 1A(B)

RESAR-3 Section 6.3.5.1 of RESAR-3 states that only "one tempera-(Q21.15) ture detector which provides heater control for the immersion heater, control room alarm and control room indication" is provided for the boron injection surge tank.

Provide the results of an analysis which addresses the effect of a single failure in this systen.

This analysis should include possible boron dilution during recirculation. Also, it is our position that the 032-lj

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CPSES/FSAR monitoring systen for the boron injection systen meet IEEE Standard 279-1971.

Discuss your intent to conply with this a

position or justify any exceptions you may have in this regard.

Response

See revised Subsection 6.3.2.2.3 RESAR-3 The description of the Emergency Safety Feature systens (Q31.16) provided in Section 7.3.1 of RESAR-3 is inconplete in that it does not provide all of the'infonnation requested in Section 7.3.1 of the Standard Format for those safety related systens, interfaces and couponents supplied by the applicant which raatch with the RESAR-3 scope systens.

f rovide all of the descriptive and design basis infonnation requested in the Standard Format for these sys tens.

In addition, provide the results of an analysis, as requested in Section 7.3.2 of the Standard Fonnat, to deaonstrate how the requirenents of the General Design Criteria and IEEE Standard 279-1971 are satisfied and the extent to which the reconnendations of applicable i

Regulatory Guide are satisfied.

Identify and justify each i

exception.

Response

See the response to QO32.17 RESAR-3 Provide analyses showing that no adverse effects will occur (Q31.17) or a discussion of such adverse effects that could occur as f

a result of power interruption to the Engineered Safety Features Actuation System at any time following the onset of a LOCA or other accident conditions.

s

Response

See Question 032.21 RESAR-3 General Design Criterion 25 requires that the protection 032-1k

CPSES/FSAR (Q31.18) systen be designed to assure that specified acceptable fuel design limits are not exceeded from an accidental i

withdrawal of a single rod control cluster assenbly (not ej ect. ion).

In the accident analysis, presented in Section 15.3-6 of RESAR, it is stated that "no single electrical or mechanical failure in the rod control system could cause th accidental withdrawal of a single rod control cluster as s embly." However, Chapter 7.0 does not describe how the design prevents such an occurrence.

Provide a detailed description of the control circuitry and discuss how the j

design meets the requirements of Criterion 25.

Response

See Question 032.29 RESAR-3 Provide a discussion which supplements those in Section (Q31.19) 7.4, 7.5, and 7.6 of RESAR-3 and which addresses the Standard Format information requirements for the safe shutdown systems, the safety-related display

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instrunentation and other safety systens and equipment outsida the RESAR-3 scope which are assumed in the RESAR-3 l

and the PSAR Chapter 15 accident analyses.

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Response

See Questioa 032.22 RESAR-3 In addition to the design features discussed in Section j

(Q31.20) 7.6.2 of RESAR-3, it is our position that the design of the RHR isolation valves satisfy the following:

(1)

The interlocks shall utilize diverse equipnent, and (2)

The interlocks shall be designed in accordance with the intent of IEEE Standard 279-1971.

3 The infonnation presented in Section 7.6.2 of RESAR-3 does 032-11 f

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CPSES/FSAR not address the requirenent for diverse equipaent and describes a degree of testability that conflicts with the requirenents of IEEE Standard 1971.

In addition it is stated that the position indications for the RHR valves differ from those for the accunulator isolation valves but these differences are not identified.

Discuss your intent to comply with the requironents that the design shall utilize diverse equipnent and shall include canplete on-line test capability without opening the isolation

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valves, or justify any exceptions.

In addition identify the differences in the position indications provided for the RHR valves canpared to the accuaualtor valves and discuss the reasons for the differences.

Response

See Question 032.27 RESAR-3 Provide the list of transients that were analyzed in deter-(5.1) mining the maximum steam system pressure transient for sizing the stean generator safety valves.

Response

See Subsection 5.2.2.2 RESAR-3 In reference to Section 5.3.4, provide Reactor Coolant

)

(5.2)

Systen Tenperature - Precent Power liap for plant with loop 1

stop valves if different from Figure 5.3.1.

Response

See Figure 4.4-21 which corresponds to RESAR-3 Figure 5.3-1:

CPSES does not have loop stop valves.

RESAR-3 Provide a discussion of the consequences of inadvertent (5.2.2) overpressurization resulting from a malfunction or operator error when the reactor coolant system is watersolid during

tartup or shutdv.1n. The discussion should include consideration of the pressure-tenperature operating 032-Im

CPSES/FSAR limitations on t'he reactor vessel to protect against brittle fracture.

In addition, discuss any design provisions that will be incorporated into the facility design to prevent overpressurization incidents that would exceed allowable pressures in this particular plant condition.

R es r'ans e See Question 212.5 RESAR-3 Discuss the ability to assure that the operational capa-(5.2.7 &

bility of the valves that are required to function in the 6.3) short and long tenn LOCA modes of ECCS operation are not impaired by potential crystallization of boric acid solutions on the valve stem due to leakage.

Appropriate methods may include the ability to detect individual valve sten leakoff or periodic operational testing of the valves.

]

Responsa See revised Subsection 6.3.2.2.12 RESAR-3 Justify the fouling factor resistance specified in Section (5.3) 5.5.2.3.1.

Correct the difference between Section 5.5.2.3.1 and Table 5.5-3 with regard to the fouling factor.

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Response

The fouling factor is discussed in Subsection 5.4.2.5.1 and is consistent with the value reported in Table 5.4-3.

4 RESAR-3 Provide pressurizeer relief and safety valve capacities (5.4) when discharging water liquid.

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Response

Liquid flow rates assuned in analysis are based on the 4

honogeneous equilibrium saturated flow model which gives the most conservative relief rate. Accident analysis

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demonstrate that water relief through the pressurizer

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032-in a

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,CPSES/FSAR valves occurs only during the feedline rupture event and the peak liquid relief rate is approximately 1 ft.3/sec at 2575 psia (see subsection 15.2.8), conpared with the honogeneous equilibrium model relief capacity of 15.3 ft.3/sec at 2575 psia.

RESAR-3 Item 6.3.11 of the " Standard Fonnat and Content of Safety (6.1)

Analysis Reports for Nuclear Power Plants" Revision 1, (October 1972) indicates the need to distinguish between true redundancy incorporated in a system and nultiple conponents.

To conplanent the SAR discussions in this regard, provide a sunmary of a systenatic core cooling functional analysis of conponents required over the conplete range of coolant pipe break inside the contai nment.

The sumary should be shown in the form of simple block diagrans beginning with the event (pipe

' break), branching out to the various possible sequences for the different size breaks, continuing through initial core cooling and ending with extended to long-term core cooling.

When conplete, the diagran should clearly identify each safety system required to function to cool the core for all coolant pipe breaks inside the containment during any plant operating state.

The attached Figure 6-1 is provided as a guide.

Response

Systen reliabiltiy of the ECCS, including a discussion of redundancy and conpliance with the single failure criteria is provided in Section 6.3.2.5.

Functioning of the various ECCS conponents for various accidents including large and small LOCAs are discussed in Subsection 6.3.3.

The actual LOCA analyses are discussed in Subsection 6.2 and 15.6.5.

Additional infonnation is provided in the response to RESAR-3 (Q15.0.1).

032-10

CPSES/FSAR RESAR-3 For each engineered safety feature identified in Question (6.2) 6.1, list the auxiliaries required for its operation.

1

Response

The supporting auxiliaries which are required to function and support the ECCS are the safeguards electrical busses, the coaponent cooling water systen, and the engineered safety features ventilation systens.

The safeguards electrical busses are required to provide electrical power to the ECCS punps and motor operated valves.

The component cooling water systen is required to provide cooling to the ECCS punps and the RHR heat exchanger (during recirculation only).

The engineered safety features ventilation systen is required to provide cooling for the ECCS punp motors.

Addition infonnation is provided in the response to RESAR-3 (Q15.0.1).

RESAR-3 For each transient and accident analyzed in Chapter 15, (15.0.1) provide the following infonnation:

i (1)

The step-by-step sequence of events fraa event initiation to the final stabilized condition.

This listing should identify each significant occurrence on a time scale, including for exanple:

flux monitor trip, insertion of control rods begin, primary coolant pressure reaches safety valve set point, safety valves open, safety valves close, containment isolation signal initiated, containnent isolated, etc.

All required operator actions should also be identified.

(2)

The entent to which normally operating plant instrunentation and controls are assuned to function.

(3)

The extent to which plant and reactor protection 032-2p

CPSES/FSAR systens are required to function.

(4)

The credit taken for the functioning of nonaally i

operating plant systems.

(5)

The operation of engineered safety systems that is required.

Response

These diagrams are given in Figures QO32.1-1 to QO32.1-24.

RESAR-3 Section 15.2.4 of RESAR-3 UNCONTROLLED BORON DILUTION, (15.0.2) analyses the effects of a dilution at power. The analysis discusses the causes of the incident, and the autonatic actions of the Reactor Protection System and the manual actions pronpted by alams and instranentation that would mitigate the consequences of the accident.

However, there is a possible situation, involving the loss of offsite power, where a dilution incident may not be as readily apparent as that described in Section 15.2.4 and where no automatic Reactor Protection System action is q

available, in order to assess the potential severity of a dilution accident after a loss of offsite power, provide the results of an analysis that assunes the anticipated equipnent configurations in nonnal use prior to the event that result in the most severe consequences. The analysis should include a dilution operation in progress with the Chanical and Volune Control System mode selector switch being in the DILUTE position (or A: TERNATE DILUTE mode).

The loss of offsite power is then assuned to occur with the minimun shutdown reactivity insertion due to control rods.

Both diesel generators start and sequence the loss of offsite 032-1q e

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CPSES/FSAR 4

power loads.

The concerns are that the charging punps again autanatically start running after being loaded to the i

diesel generators and fran electrical schenatics of control j

circutis for the reactor makeup water punps, that the i

reactor makeup water punps would also again automatically start with the mode selector switch in DILUTE.

Therefore, a dilution of the~ Reactor Coolant Systen is again in f

progress which could potentially result in a return to i

critical.

l If the reactor makeup water batch integrator is assuned to malfunction by not automatically cutting off flow at the 1

pre-selected value, provide the time available for manual action before the total shutdown margin is lost due to this l

dilution.

If operator action is to be pronpted by alanns, describe the features that will alert the operator to this

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specific action at a time when alarms from many plant systens are occurring simultanelously.

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Response

This question will be answered with the responses to round one questions.

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i ABBREVIATIONS USED:

AFWS - AUXlLIARY FEEDWATER SYSTEM ECCS - EMERGENCY CORE COOLING SYSTEM CVCS - CHEMICAL AND VOLUME CONTROL HL

- HOT LEG l

SYSTEM CL

- COLD LEG ESFAS - ENGINEERED SAFETY FEATURES CCWS - COMPONENT COOLING WATER ACTUATION SYSTEM SYSTEM l

FW

- FEEDWATER RCS - REACTOR COOLANT SYSTEM RTS

- REACTOR TRIP SYSTEM SWS - SERVICE WATER SYSTEM SIS

- SAFETY INJECTION SYSTEM HPI - HIGH PRESSURE INJECTION SI

- SAFCTY INJECTION LPI - LOW PRESSURE INJECTION RT

- REACTOR TRIP C1

- CONTAINMENT ISOLATION CS

- CONTAINMENT SPRAY SG

- STEAM GENERATOR

+

NOTES:

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1.

FOR TRIP INITI ATION AND SAFETY SYSTEM ACTUATION, MULTIPLE SIGWALS ARE SHOWN BUT ONLY A SINGLE SIGNAL IS REQUIRED.

THE OTHER SIGNALS j

ARE BACKUPS.

2.

NO TIMING' SEQUENCE IS IMPLIED BY POSITION OF VARIOUS BRANCHES.

REFER TO EVENT TIMING SEQUENCES PRESENTED IN TABULAR FORM IN PERTINENT ACCIDENT ANALYSIS SECTION OF CHAPTER 15.0 0F THE FSAR.

DI AGRAM SYHBOLS:

J h

j

(

) - EVENT TITLE

- BRANCH POINT FOR DIFFERENT PLANT CONDITIONS

- SAFETY SYSTEM

- SAFETY ACTION (S

- SYSTEM REQUIRED TO MEET SINGLE-FAILURE CRITERIA

(

- MANUAL ACTION REQUIRED DURING SYSTEM OPERATION COMANCHE PEAK S.E.S.

i FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 ABBREVIATIONS FIGURE QO32.1-0 l

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(TLtDaA1(R Sf3 FEM MAlf0MCil0NS ACESSIVL H[ AI RtH0d AL trE TO

~~

F.G d O V q, s uo 10 Al E bif "i ll P:14 T Lua

.q'i k ui g i A;( d 4 '.c I /2 HIGH NE UIROM F LUt P0wt R R A%f 2l%

• i4=

' '. 'j i P:.14 k t%s 2 /.

P>iR < PIO./ \\ PGalR > PIO V

Ovf RP0af R 2/4 0'l'***

2!4 RTS RT5 OdERitHPERATO'f 2 /4 e

\\

Q,/

Q,F/

3 f

J RE ACION kEACTO' IRIP TinP Bif A RI Ri 8RE Anf R3

/%

/~'%

\\([/

\\((/

C0h!W R0its (ONIN0L R0J3 GDAiliY GRAflif I N ' E R ! ! 0h InifRil0h PASilv!

Pa5 sty!

(0ml#3L d OD (09ts0L NOJ REACilittY

1. [ACilvily (OM!V T.

(04 T70L f 0. Pw s,,

1 i

\\

COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 EXCESSIVE HEAT REMOVAL DUE TO FEE 0 WATER SYSTEMS MALFUNCTION FIGURE QO32.1-1

e

[1CF.SSIVE LOAD INCREASE

- (FULL POWER )

0 (E P P0a EW 2/4 O tE R IEMPE R A TURE 2 /4 PO*fR R1 0E HIGrf NEUIRON 2/4 Flut

-+

R IS PRESSURllER STEAM SAFETY GENERATOR V ALVE S SAFETY VALVES

/

'\\

F i PASSIVE PASSIVE gS W

/

REACTOR IalP BRE AKERS SAFETY VALVES SAiEIY Y LVES OFEM IO REllEVE OPEN TO R(t IEVE r'

RCS PRESSURE SECONDARY SOTEM PRESSURE l5 p j

\\.

/

0% 1. THIS DI AGRN AWES TO 30TH CONTROL RODS MANUAL AND AUTCHATic CONIROL GRAVlif b*

IN3ERil0M

2. F O'i Till S TR ANS I E N T. R E AC T OR PASSIVE PROTECil0M SYSTEM FUNCTIONS ARE ASSUMED TO BE OPER ATIVE, BUT A REACTOR TRIP IS NOT E X PE C TED.

CONTROL R00 RE/CIIVLTY CUNIROL i

i COMANCHE PE AK S.E.S.

RNAL SAFETY ANALYSIS REPORT UNITS 1 and 2 EXCESSIVE LOAD INCREASE FIGURE QO32.1-2

Hisy NEL'? Reg rLgx CEPRE33'J4!2 ATiew of

" AIM S'E8" SVIIEM p3Eg eA*SE 7/4 OVERP0mER 214

$1 SIG4AL Lfd PRE 35 021T; 4 2N l

M igd ?-C04' A 14MT W' 7/3 St 5 ' C 4& L i

1 III PR L 55 t;k t

)

m s.,

2/3 NE s5UR L p

PGE5$'.9E :n 1 LCOP l

s/G p;p sig 2/3 j L** S'E)ML t hE 7/3 I

']

Luh! Ainsn! HEnunt g$r g$

ENI

/

,E.,tt i n Ay

'l 3

s j,#

LO*' $TE AM Lig

"'D kIGAT' E S'E AS 0/3 I/G I

PR ESSURE ;M OkE N

REACTCR LGCP 2 /3 s' 8 f)

PfiS5URE WA!E 1% 1 LCSP

$ p\\

,'g }%

-)

( > Pil) r s

%. s

~

TRLP 38EACRS MAM' AL i 12 J

g l

STE AM Liu clu I

l

+!PLATICR F4 V Alv E3 l$0LAijgg

($

p)i L

/

aS Fl NY CC1'iOL RCDS GR A d lTY tgaECTIC4 0F 14 SERT?C4 33 RATE 0 MATER CLCSE FA5!

ct0cE MA f M FW PA5$ 1EE TO P8En ENT ACT;E3 $iEA4 ISOLAT I Ch U Alh E$

REiuRu TO UNE i!OLATIOu CRITtCAll!Y

• k ALVE5 I' SIS"AL l

CON'7CL R00 r$FAS RE ACT iv 1T Y CC9:ROL LOW-LCW S/S L! VEL 2/3 14 6 3/G

's e

'C*ERATOR TERuinATES SArEiv iMJEC' 01 l

j pq FLOW TO Limr 7 RCS PRE $$URE Ag3 p3 4,'A L e /2 I

I FRE tSue t2ER LE VE L.

l}

SEE T A!iE 7.5 ! FOR is0 :C A T OR!

,, g AND iiECORJER$ A. A L SLE TO 'hE P v*5)

~,

CPER A!Og F0,Lew s3 rt EgEsi.

S!G (!UE3!hE OR HEN t

OE LlER AUIILI Ally FEE 0aATER 70 C04!RCL CORE

• E AT REuos AL COMANCHE PEAK S E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 DEPRESSURIZATION OF MAIN STEAM SYSTEM FIGUREQO32.1-3

LOSS OF EXTERNAL ELECTRICAL LOAD j

FULL P0 DER RTS HIGH PRESSURIZER 2/4 PRESSURE OVERfEMPERATU M 2/4 e %

iS S \\

% s' MIGH PRESS:JWI/Fo 2/3 PR E S SUR ilE R SIEAM WATER LEVf L SAFETY GE NE R AIOR RE ACTN VALVES SAFETY VALVES pp BREAKERS PASSIVE PASSIVE

' 'N t S

\\

}

s CONikOL RODS S AFETY V ALVES SAFETY VALVES GRAVITY OPE N 10 REllEVE OPEN 10 REllEVE INSERIl0N RCS FRESSURE SECONDARY SY3 TEM PRESSURE pgg3;g CONTROL R00 REACTIVITY CONTROL L0d-LON S/G LEVEL IN I S/G 2/4 ESFAi MANUAL l/2

[ %

gS F1

/

s l

l AF#S

'f 09 CASE wMERE TURBINE TRIP OCCURS,

/ 5 i S Fi A RE ACTOR TRIP SIGNAL ON TURBINE TRIP l$ ANTICIPATED (FOR P0 DER > P7) s IURBINE ute SIGNAL DUE TO:

2/3 LOW TRIP FLUID DEllVER PRESSURE AUXILIARY FEED #ATER TO 4/4 TURSINE STOP g

VALVE CLOSURE HE AT REM 0'/ AL COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 LOSS OF EXTERNAL LOAD FIGURE QO32.1-4

LO35 Or 0:rsi![ Pc*-ta tc 3T A 10s AiI LI A:s IES (SLA # 0UI)

F UL L PC*E 9 LOW-L0w 5.G. WATER LEVEL 2/4 IN A4Y LOOP l

l l

l Auf lC I Pi?E ? R.T. 0*l TCR3!NE g rS ppg 3SLgjyta

$*EAN ESFAS TRi?

S AFE ry GE*i4ATC8 LC -L0w 5.5. LE.E L 2tc

,[

AL;ts SITETY.ALLES

',I 5!U LOW eE A!!OR COOLA%T

['g T LO.' I N I LOOP 2/3 3

f Pa$$l 'E P153!VE F

j j

1 I

T I

l LOL$ Or Or r51TE PO*ER TRIP SArETY v1LWE5 MA*uai. :/2 SRE A GS enru to

$ELOh:ARY RELLE4E RCS SYS4M I

1. 1. ES D E f

g PR E S S A '

2/4 t 0=.LO. LE:! L ! N S f g

j

, s/c t', 3;w OR;.Es

_,,\\

PLos ;

/

~

S\\--'/

004'ROL RODS DELIVER actlLIARY PA SS I'4 E FEE 0aA!ER TO l

COCROL CORE f

< A' af v0V AL CONTFOL R02

• EACT19TV C ChiY Qs i

COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 LOSS OF 0FFSITE POWER TO STATIONAUXILIARIES(BLACK 0UT)

FIGURE Q032.1-5

~

.c...

. = - - -

- ~ -

4 A LOSS OF 40NAL FEE 0 DATER Y

- FL'LL POWER ON ERTEMPER ATbRE 2/4 LCW. LOW 3.G. WATER LEVEL 2/4 PRESS'RIZER SIEAM LOW.LD'n S.G. LEVEL 2/4 ESFAS J

in AMf LCCP RTS SAFETY CENERATOR m i SM '

VALVES SAFETY b1LVES

,-~s

~

1S F 3 s

3 F 3 PASa lif.

PASS 1wE LOSS OF OFF5 tTE PC.TR s,,f f

NA!tJA L. 1/2 AFa5 '

TRIP SAFETY VALVES f L

'L L*

BREAKERS SAFETY VALVES OPEN TO GEttEVE 2/4 S/G (TURBihE OPEN TA RELIE(E SECCscARY SYSTEM ORivEN PUMPS)

RCS PRESSURE PRESSL'RE r\\

IS F) i

%/

IS f t

%,,/

CONTROL RODS DELMEE GRAvlTY Aut!LIARY INSERTICN FEE 0 WATER TO PASSIVE CONTROL CCRE PEAT REMOWAL

)

CONTROL RCD REACTIVliY CC4 TROL COMANCHE PEAK S.E.S.

FINAL SAFETk ANALYSIS REPORT UNITS 1 and 2 LOSS OF NORMAL FEEDWATER FIGURE Q032.1-6

.i

_ _., __ i_j

MAJOR RUP'U1E Of A MA l m FEEDaA*ER Lt ni

- FULL PC ER h ! Git PRIS$U9 TIER PRE $$uRE Of 4 J RTs l

S Ar!T V !*J!cTits sts>AL

% f Agu wE 0/3 OkERTEMPERANRE 2!4 PFE$Mi IM i LOOP f-]')

LO"-LU" IIS LE VE L l# A" SIG 3

, g, A WT 2/3 LOW-Lee 3 /J LEVEL t w AMV g

IN II" P4 E35 J E E5fA5 E$FAS (bOIOR PN$} 2/R LCw-L0m LEVEL tu fic $/3

$! SIGNAL REgCTLR

-'g (ICRSIME PyMP)

Trip g

8REA'E'S

/, 7}

(3

/

64A v ;t t 5- /

tc55 cr errsivE PMR

( Af*3

{

f PS E$ $1;R IZE R

\\-

l SAFETY 5'S t ttvEs C0h!HOL R003 PAsil d E

$8AviTV

,,'h

,,h

[

14$ERTt04 I

Y)

t

\\s PAS $ t hF

~-/

DElldt mECT !0 APES 5 AF ET Y v tL v Ei

iLlk

• • II" 'C0 "C5 CON'20L R00 CPD TO FEE'F ATER TO

'O PEE d k:

R E AC T it I TV REUE4E "Y'

CCM'ROL RC$ PetE$$ GEE dE AT REaa05 A'.

TC PM R -

' CPER ATOP ILR%$ Of f MICH

• E A3 $1 'E TY f MJECT ICM PLMP5 5)SSECuEu? 70 REthERv er LnEL t o TFE in! ACT $lG's 5!E TABLE 7.54 EOR 4

to:CATOR$ AMO RECOROERS AVA!LAhE TO Tel OPER A'OR f 0LECaln3 inE EJEh!,

COMANCHE PE AK S.E.S.

FINAL SAFETY A.NALYSIS REPORT UNITS 1 and 2 I

MAJOR RUPTURE OF A MAIN FEEDWATER LINE FIGURE 0032.1-7

r%,

LOS5 Or FCRCE3 RE ACTCR COOLA*T F LC=

PCa!R < PT POWE R > P7 V

NO PRO!ECT :Ch PAE' til LCSS

/\\

TOT AL (003 V

- SINCLE RC PUu? LCCRE3 RC'CR, $MF? BFEAR PC ER > P6 power < PS (F lLL PCsEE) t 2/3 LOW R(3

{

(;g;g g y3(!4CE pfg liv RCS FLCW IN i LCCP 2/3

( tg, g y y (999 R!s RT5 RTS RTS

'"2I'f EY"C' 2/3 lcm ROS t Lt> tn 2f4 LOOP

/N

/~N

/~'\\

(nw ots Flos

/'N 3

l5 gs ry gs Fj r1

($

F L N CME LCOP s,

(,/

N/

sj klaC T ON p g ACTC9 EEACTOR RE ACT0w

,,(g$ ug ;7(,

IIIP TRIP TR p TGIP gy nargy:qg 94[twf SS j,E A'f R S g

gin og f

Pas s n t f-s j

($,F I (3

F (5.Fj

($

F

/

/

s-s.

l l

(C4TFOL R005 (CgTROL R033 CONTROL R005 CONTROL RC05

$AFETV WttvES ggg.; g, G9 Ai 11

• Grid :

GR Af I T ~f 0 u 0 EE INSE R T !Cn 1 % 5 E U 1 **

. 45T R T IC4

,,$g gy :;,

P 45S i h PAS S h E

• iSS i E PA55 t sE v

I

\\

\\

CONTROL #C0 C04'ROL 8 00 ConTRO *03 ccstret ROD g

RC ACI IV I T Y E E AC T ! d !TY REAC',VlTV RE AC T ;V ITY C01!ROL CON'?0L CC%i:0L CCnTROL COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 LOSS OF FORCED REACTOR COOLANT FLOW l

FIGURE Q032.1-8 et

UNCCNTROLLE3 000 CLUSTER CC4740.

A$$EMPLY $1%K Wl'FOR&*AL SOUR:E RAMGE HION h20'R0h Flux,t/2 SUSCR t TICAL

/\\

AT PCwf R INTERMED ATE RANGE H!

• NEUTRC4 i LUX. }l2 gE'. RANCE N!OM ME U'RO N FLJX. L0a. 2/4 PCwER PANM etGH %EUTecu_

RESCTC9 FLtX. H1St. 2/4 RFACTCR PRLs J #2ER S/G U#

TRIP

$4ft'f SAFETV 5)$ TEM P0wf? Ra%3E HtCH w?O'Romk SYS'E4

.ALkES VALSES 1

Illl EAIE 2I#

PiS : s E PA53liE eis f--s (SIFl (S r,

\\

/

2/a PSERRA6E wigs

\\M hEUT9Ch i LU X EEAC'02 REAC'C3 SAFETY VAltES TRIP SAFETY kALVES c p[ g *c g,LtEy[

73 %[qr[wtagiggt Ul?

n;[gv[g; BREAFER$

C PE N TO PEltE E

$ggg g, ey37gg k:S 8RESSLCE pagg3;g lgh

N 2',

c Esp;aER

(-

I $

Fj

%.s 2'4

• OH PRESSvai?!D I

l i CO O OL RCOS CONTPOL ' 005 pq ss,;,

G'AWtTY caraiTy CONTE 0L IM3ERT10m 2l3 wisy pgEssgpi:ER PassigE

.1TER f.E L P!SSIsE 0 "' E00 CONTRCL ROD

" A0 " O REACTIV!!Y Uk CONTROL COMANCHE PE AK S.E.S.

F!NAL SAFETY ANALYSIS REPORT UNITS 1 and 2 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL

4 T

1 DROPPED k00 CLUSTER CONTROL ASS EMBLY

• POWER P,ANGE HIGH NEGATIVE RTS NEUTRON FLUX R ATE 2/4 p..

\\

I S F

L l I

REACTOR TRIP BREAKERS t-1

~'N

/

IS F)

\\_ s 6

CONTR01 RODS GRAVifY I NSEP,T ION PASSIVE i

CONTRCL R0D REACTIVITY CONTROL i

• TRIP SEQUENCE OCCURS ONLY IF R00 WORTH IS GREATER THAN MINIMUM REQUIRED TO TRIGGER THE FLUX RATE TRIP, OR IF AN ENTIRE RCCA BANK DROPS.

FOR OTHER RCD fl0RTH, OR FOR MISALIGNMENT, NO PROTECTION REQUIRED.

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2

' DROPPED R0D CLUSTER CONTROL ASSEMBLY

), FIGJnE QO32.1-10 m

e av-pgWy-r T

W-

--'-e---w'e

• Wr-'*-r*-*N+--

w e"N1't

• -9--"-

v

'f'"H

^t-'Mw'

Mwr' W

G'T""#'

'D

I SINGLE ROD CLUSTER CONTROL ASSEM3LY WITHDRAWAL AT FULL POWER i

OVERTEMPERATURE 2/4 REACTOR

)

TRIP SYSTEM

(,I LS F

\\

REACTOR TRIP BREAKERS

,c

\\ S F j

^

mj i

CONTROL RODS GRAVITY PASSIVE CONTROL R0D REACTIVITY CONTROL COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 SINGLE R0D CLUSTER CONTROL l

ASSEMBLY WITHDRAWAL AT AT FULL _20HER FIGURE 0032.1-11 a

i

~

I STARTUP OF AN INACTIVE REACTOR COOLANT PUMP J MAXIMUM PERMISSIBLE POWER FOR OPERATION WITH I LOOP OUT OF i

SERVICE i

HIGH NEUTRON FLUX POWER 2/ti RTS RANGETRIP(HIGH)

/

'\\

lS FJ

\\_./

b REACTOR TRIP BREAKERS i

i tS F \\/

i N -

i CONTROL RODS GRAVITY INSERTION PASSIVE CONTROL ROD

^

REACTIVITY CONTROL COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 1

STARTUP OF AN INACTIVE REACTOR COOLANT LOOP FIGURE 0032.1-12 J-

..u.

m.

...m

.m m...m_

m m 4 m m.

. _. _.~.

m.. -. _

a-.

t 9

RcRen O tLUTICN SUSCRITICAL AT PC*e1R REFUEL.t us not. COLD 5*uvs==. s w M V

43?cmA2.C CCM'ROL MAmusi CON'ROL V

I R00.sEtrio.-

I L ist. f ALAm*ts

(

ROD laSERiten AJDISLE. C0' nT R ATE CVC1 L:Wii ALAEW e

$7EP togn'ER C7C$

11FE rv CEhE2ATC#

FRN 8F3. DETECTmsk @

h 80D Pe t f a ca W ALw E3 1AFETY salbE3 I

PA5 s t hE Pass a rE eraCTOR TRtP ALAPp I

I TERM t h ATE TERMINATE 3ArE!v valves DiLUT tDs 08'2I30" SAFEn vams Mu to REL:ErE TERNimi!E CPEu TO RELIEvf 3ECce:ARY Sv5ffw 3,gg7 gn,,

203 PRES 3cRE PREggpE (NG'E 4) 2 /4 F*s ER R ANCE u f Cd m

CvC1 ntuTRCn f tur RTS 2fs DVEarEMPERA?cRE

\\

AUDIBLE COUNY RATE FROM

!5 F

(s) 2fg CVERP0w!R 91 DETECYORS 3

etGd 50L9CE R ANGE Nfe'RC4 ppg

, qg,

t FLUS ALARw TEdu t KATE PWL a$W t.

TR IP g,ty7 i gg RRE A KERS I

(t0TE A)

/*N-LD E L 58 81

%s

?D

• L0w AND Low-LCw LiktT Com'50L RODS

• • he' 5Af ETV $f $ TEM 6EAdlTF I NTERT (CM NOTE &. CLCSE MANEUP WATit ConfeCL D ALEES.

CLCSE CHA96!NG LlkE Courtet VALVE 5.

PAS

• tVE TERwinATE CHARS #S PUNP FL0w l

BOR ATE tF MECESSART I

CCMTROL ROD REACT 1v ITY j}

ces?s-ot t

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 1

i B0PON DILUTION FIGURE 0032.1-13

=-

4 a.--

=%.;

m a_

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h

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. ~,,. - _. _ - _ ~ -,, -..,, __r._,.__,_.

J INADVERTENT ECCS OPERATION AT POWER FULL POWER LOW PRESSURIZER PRESSURE 2/tl MANUAL. i/2 l-r s

I\\_.p }

3

/

REACTOR TRIP j

BRE A KERS

/ N IS F)

\\

./

CONTROL RODS G:t AV ITY INSERTION 1

CONTROL R0D I

REACTIVITY CONTROL l

COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 IllADVERTEt!T ECCS OPERATI0ll

.AT POWER FIGURE Q032.1-15

m-ACCIDENTAL DEPRESSURIZATION OF REACTOR COOLANT SYSTE4 hFULLPO'ER w

OVERTEMPERATURE 2/4 l

l LO PRESSURIZER PRESSURE 2/4 2/4l RTS ESFAS l MANUAL 1/2 PRESSURIZER LOW PRESSURE'

,s F,

($

F1 g

j

%./

s s SIS REACTOR TRIP BREAKERS tS F1

/"N

%./

(S F)

INJECT ECRATED WATER INTO RCS TO PRE.' NT CONTROL RODS RETURN TO PCwER ASD CMVliy MAINTAIN CORE C00Lihs INSERTION PASSIVE l

CONTROL RCD REACTIVITY CONTROL COMANCHE PE AK S.E.S.

l FINAL SAFETY ANA. LYSIS REPORT I

UNITS 1 and 2 ACCIDENTAL DEPRESSURIZATION OF REACTOR COOLANT SYSTEM l

F:GURE Q0.32.1-16

w.

e e

1 1

SSchtf f ut'(0 s tP

• um? & e=f S it.ad Ci nt 4 A *C4

? ?$1 I

i CP( ta ? 0# (97) ~

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J%y' FINAL SAFETY ANALYSIS REPORT '

CE>Laa'0' UNITS 1 and 2 s

STEAM GENERATOR TUBE RUPTURE FIGURE Q032.1_-1_7

LOCA REQu! RED FCR BREAK 5 tieALLER TMA g i FT2 CP!qiTOR (01) -

p (2/4) L0 PRESSURtIER P -

(2I") "I'3 CE"IAI"ME"I #

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CONT. 150 g;gggggtgyggq SREAKERS ISOLAT ION

1. A 3YSIEM LPt

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CCNT. PRES.

Ccg3 CONTROL POD CONTROL ECCS HL HEAT REMOv AL RE A0T W I TT SEC0%3ARY REC ROULATI0g As0 ACTiv TTY C%!ROL HE!.T REv0ML CONTROL

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' CPER ATOR ACTION IMOLUDES WALVE ALIGkMENT AND PLNP CCNTROL REQUIRCO TOR RECIRCULA!!CN ggppgg7 "CPERATOR ALIGM3 SPRAY PUMP SUCTf 0N FCR PECf RCULATICM T0 SYSTEMS CORE C00Lin3 C0hTAthMCNT RECIRCULATION SLMPS OM LC-LC RWST LEvf L AL ARM CPE R Ai ! CNA L

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COMANCHE PEAK S.E.S.

FOOTn0TE: SEE TABLE 7.5-1 FOR IM0f CATC#3 ANO RECORDERS AVAILA8tE TO THE OPERATCR FOLLCWING EVEh?

AL SAFETY A.NALYSIS REPORT UNITS 1 and 2 LOSS OF COOLANT ACCIDENT FIGURE QO32.1-18

CVCS LETDOWN LINE RUPTURE sr REACTOR COOLANT ACTIVITY RELEASE TO SAFEGUARDS BLDG ATMOSPHERE UNFILTERED RELEASE TO ATMOSPHERE SAFEGUARDS BLDG DUCT RADIATION MONITOR s,

OPERATOR @ CLOSES LETDOWN ISOLATION VAL'lE AFTER RECEIPT OF THE ALARM WITHIN 30 MINUTES /

l i

COMANCHE PE AK S.E.S.

j I FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 CVCS LETDOWN LINE RUPTURE FKRJRE 0032.1-19

GWPS GAS DECAY TANK RUPTURE F

RELEASE TO BUILDIrlG ATMOSPHERE UflFILTERED RELEASE TO ATMOSPHERE VEtlTILATION VENT RADIATI0ll MONITOR

.c OPERATOR TERMIt!ATES PROCESS GAS IllFLUEllT ISOLATION TIME 2 iiRS COMANCHE PE AK S E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 GWPS GAS DECAY TANK RUPTURE FKiURE Q032.1-20

=-

'l FLOOR DRAlft TANK FAILURE (30,000 GAL)

LIQUID

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i RELEASE TO GROUND RELEASE TO BUILDING ATf40 SPHERE 4

ENTERS UllFILTERED RELEASE GROUNDWATER TO i

ATMOSPHERE

• See Section 2.4.13 COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 FLOOR DRAIN TANK FAILURE FIGURE Q032.1-21

..... --...~...

FUEL HAtIDLING ACCIDENT Ifl FUEL BUILDING v

GASES RELEASED FROM FUEL POOL WATER

,s

( SF v,-

V FILTERED BY PRIMARY PLANT VENTILATION SYSTEM 10DIllE ADSORBER TO ATMOSPHERE v

RADIOLOGICAL CONTROL w

COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 FUEL HANDLING ACCIDENT IN FUEL BUILDING FIGURE QO32.1-22

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FUEL. HAf!DLING ACCIDENT INSIDE C0flTAINMENT s.

ACTIVITY RELEASED FROM REFUEllflG CAVITY l

Y i

FILTERED BY C0flTAlfiMEllT AIR CLEANUP UNITS TO ATMOSPHERE Y

RADIOLOGICAL CONTROL I

4 COMANCHE PE AK S.E.S.

RNAL SAFETY ANALYSIS REPORT UNITS 1 and 2 FUEL HANDLIflG ACCIDENT INSIDE CONTAINMEllT FIGURE 0032.1-23

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1 l

SPENT FUEL CASK DROP l

(FROM HT 30 FT) i 8(

NO RELEASE; NO CASK DAMAGE 3

4 4

a b

COMANCHE PE AK S.E.S.

I FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 SPENT FUEL CAS'K DROP ACCIDENT

^

FIGURE Q032.1-24

. -.... -.. ~ -

CPSES/FSAR C

To prevent cold spots and stratification within the tank during' normal operaticn, the contents of the boron injection tank are continuously recirculated with the boron injection surge tank via a boron injection recirculation pump.

The boron injection tank incorporates a sparger type inlet which distributes the incoming boric acid in a 360 degree fan as it enters the tank.

This prevents channeling and also ensures homogeneity of the boric acid solution.

This recirculation path is automatically isolated on receipt of an "S" signal.

Redundant tank heaters and line heat tracing are provided to ensure that the solution will be maintained at a temperature in excess of its solubility limit (135 F at a nominal 12 percent concentration of 21,000 ppm boron).

6.3.2.2.3 Boron Injection Surge Tank

(

The boron injection surge tank provides surge capacity for the boron injection tank recirculation loop.

The boron injection surge tank con-tains the same concentration of boric acid as the boron injection tank during normal plant operation.

The recirculation lines to and frcm the boron injection surge tank are automatically closed and the boron injec-tion recirculation pumps stopped by the safety injection "S" signal.

An inmersion heater is provided to keep the temperature of the solution high enough to prevent precipitation of the boric acid.

P,ESAPM INSERT B l(Q 31./5) 6.3.2.2.4 kc:idual' Haat Removal Pumps In the event of a LOCA the y esidual heat removal pumps are started automatically on receipt o* an "S" signal. The resi-dual heat removal pumps deliver water to the RCS from the refueling water storage tank during the injection phase and from the containment sump during the recirculation phase.

Each residual heat removal pump is a single stage vertical position centrifugal pump.

[

6.3-7

CPSES/FSAR A minimum flow bypass line is provided for the pumps to recirculate and return the pump discharge fluid to the pump suction should these pumps be started with their normal flow paths blocked. Once flow is established to the RCS, the bypass line is automatically closed.

This line prevents deadheading of the pumps and permits pump testing during normal operation.

The residual heat removal pumps are discussed further in Section 5.4.7.

A pump perfonnance curve is given in Figure 6.3-3.

6.3.2.2.5 Centrifugal Charging Pumps In the event of an accident the charging pumps are started automatically on receipt of an "S" signal and are automatically aligned to take suc-tion from the refueling water storage tank during injection.

During recirculation, suction is provided from the residual heat removal pump discharge.

These pumps deliver flow through the boron injection tank to the RCS at Each centrifugal charging pump is a mul-the prevailing RCS pressure.

tistage diffuser design, barrel-type casing with vertical suction and discharge nozzles.

A minimum flow bypass line is provided on each pump discharge to re-circulate flow to the pump suction af ter cooling via the seal water heat exchanger during normal plant operation. The minimum flow bypass line contains two valves in series which close on receipt of the safety injection "S" signal.

This signal also closes the valves to isolate the normal charging line and volume control tank and opens the charging pump / refueling water storage tank suction valves to align the high head portion of the ECCS for injection.

The charging pumps may be tested A pump perfor-during power operation vic the mini:aum ficw bypass line.

mance curve is given in Figure 6.3-4.

bI 6.3-8

CPSES/FSAR

.i 6.3.2.2.11 Accunulator Motor Operated Valve Controls

{

As part of the plant shutdown adninistrative procedures, the operator is required to close these valves.

This crevents a loss of accumulator water inventory to the RCS and is done shortly after the RCS has been depressurized below the safety injection unblock setpoint. The redun-dant pressure and level alarms on each accumulator would remind the operator to close these valves, if any were inadvertently lef t open.

Control power is disconnected to these valves after closure via Control Rocm switches.

i During plant startup, the operator is instructed via procedures to energize and open these valves when the RCS pressure reaches the safety injection setpoint.

Monitor lights in conjunction with an audible alarm will alert the operator should any of these valves be left inad-vertently closed once the RCS pressure increases beyond the safety

)

injection unblock setpoint.

i k

The accumulator isolation valves are not required to move during power operation or in a post accident situation.

For a discussion of limiting conditions for operation and surveillance requirements of these valves, refer to Section 3/4 5.1 of the Technical Specifications.

For further discussions of the instrumentation associated with these valves refer to Sections 6.3.5, 7.3.1 and 7.6.4.

6.3.2.2.12 Motor Operated Valves and Controls Remotely operated valves for the injection mode which are under manual control (i.e., valves which normally are in their ready position and do not require a safety injection signal) have their positions indicated on a common portion of the control board.

If a canponent is out of its proper position, its monitor light will indicate this on the control 6.3-15

CPSES/FSAR panel.

At any time during operation when one of these valves is not in

(

the ready position for injection, this condition is shown visually on the board, and an audible alarm is sounded in the Control Room.

t i

The ECCS delivery lag times are given in Chapter 15.

The accumulator injection time varies as the size of the assumed break varies since the RCS pressure drop will vary proportionately to the break size.

j Spurious movement of a motor operated valve due to an electrical fault in the motor actuation circuitry, coincident with a LOCA has been analyzed and found to be a very low probability event.

However, to comply with the NRC's present position on this issue, the applicant has comitted to compliance with BTP-EICSB-18.

Compliance is accomplished by providing a control board control power cut-off switch for each i

valve whose spurious movement could result in degraded ECCS performance.

l The applicant, nevertheless, reserves the right to retract this commit-ment in light of further analysis being generically conducted by Westinghouse.

(

j Table 6.3-3 is a listing of motor operated isolation valves in the ECCS showing interlocks, automatic features and position indications.

RESn?,-3

.E~NSERT A

( g g.7 )

asz

,3

< 6.3.2.3 Applicable Codes and Classifications 1

Applicable industry codes and classifications for ECCS components are

]

discussed in Section 3.2.

6.3.2.4 Materials Specifications and Compatibility 1

Materials employed for components of the ECCS are given in Table 6.3-4.

Materials are selected to meet the applicable material requirements of the codes in Table 3.2-2-and the following additional requirements:

1.

All parts of components in contact with borated water are fabricated of or clad with austenitic stainless steel or equivalent corrosion

{

resistant material.

6.3-16

INSERT A Periodic visual inspection and operability testing of the motor operated valves in the ECCS insures that there is no potential for impairment of valve operability due to boric acid crystallization which could result from valve stem leakage.

IflSERT B If the boron injection surge tank heater is inoperable, suf-ficient design capabilities of the 12 weight percent boric acid system are available to prevent precipitation within the boron injection surge tank.

The boron injection recirculation pumps are designed to con-tinuously circulate the concentrated boric through the 12 weight percent boric acid system in order to maintain a uniform temperature in the system and to assure a uniform concentration of boric acid throughout the system.

Heat is provided by redundant strip heaters located on piping, valves and pumps.

In addition, the operating recirculating pump provides a heat input.

Finally, there are two temperature indicator-controllers located in the boron injection tank. They indicate locally the temperature of the concentrated boric acid and control the electrical strip heaters on the boron injection tank to maintain a predetermined fluid temperature. They are set to alarm if the fluid temperature deviates from this pre-determined temperature by more than a set amount.

P b

4 4

t

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