ML14281A555

Susquehanna - Post Exam Comments, Resolution, and Technical References (Folder 1)

ML14281A555
Person / Time
Site:	Susquehanna
Issue date:	09/09/2014
From:	Operations Branch I
To:	Susquehanna
Shared Package
ML14079A115	List: IR 05000387/2014301 ML14083A036 ML14212A264 ML14220A037 ML14281A555 ML14282A459 ML14282A527 ML14288A319 ML14288A363 ML14288A365 ML15146A419
References
TAC U01896
Download: ML14281A555 (93)

Text

NRC Resolution of facility comments.

077: Facility Comment: Accept a second correct answer as a subset of the original correct answer. NRC Resolution:

The facility submittal provides a clea*r walkthrough on graphs as to why their proposed second correct answer is a subset of the original answer. Accept two correct answers as proposed.

078: Facility Comment: This question was modified to make it SRO level and a stem focus issue resulted.

The original correct answer depends on whether or not "extremely high vibration" exists at ON-192-002 step 3.4. This term is not actually defined anyplace.

This makes the choice between the original answer and the proposed second answer a matter of CRS judgement.

The facility comment is more involved, concerning operator expected response to actions NRC Resolution:

The NRC identified the definition issue during review of the facility comment and considered this to be adequate justification for accepting the facility resolution Accept two correct answers as proposed.

080: Facility comment: The question stem presents a combined electric, then hydraulic A TWS. All physically available (given the stem conditions) methods of rod insertion are available with no preferred method. Accept two correct answers. NRC Resolution:

Review of E0-1 00-113 sh 2 indicates that there are preferred methods for an uncomplicated electric or hydraulic A TWS, but all methods are "on the table" for the situation presented in the stem. Review by 6 NRC examiners did not result in one clear correct answer. Accept two correct answers as proposed.

SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS Following the administration of the LOC-26 written examination on August 21, 2014, SSES conducted an examination analysis in accordance with NUREG-1021 ES-403 D.3.a and reviewed the results of the preliminary grading and examination analysis with the applicants.

No changes were made during the administration of the exam. From the exam analysis and post-exam review with the applicants SSES has identified three written exam questions (SRO 77, SRO 78, and SRO 80) where a change in the answer key is requested.

During the exam there were no questions asked by the applicants regarding these three questions.

The requested changes to the questions are founded in procedure requirements and the bases are described in detail below. The complete examination analysis is included as part of the submittal package. Page 1 of 13 SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS SRO QUESTION 77 Refer to the figure on the following page when answering this question.

Unit 1 is operating at rated power with main generator operation as shown. Transient grid conditions result in oscillations in generator reactive load. Main generator reactive load begins to oscillate between 200 and 300 MVAR. Annunciator GEN VOLT REG AUTO TO MAN SETPOINT UNBALANCED (AR-106-C09) is in alarm. Annunciator GENERATOR FIELD OVERVOL TAGE (AR-1 06-A06) remains clear. Which one of the following describes the appropriate actions to direct in response to the conditions represented by the process computer display? A. B. C. D. Verify the Auto Voltage Regulator automatically maintains Generator Field current < 6000 amps Adjust HC-1 0002, MAN VOLT REG ADJUST, as necessary to clear AR-1 06-C09 Immediately transfer to the Manual Voltage Regulator Lower HC-1 0002, MAN VOLT REG ADJUST, until generator reactive load is < 150 MVAR Reduce core power per the CRC instructions to lower generator load to restore positive margin to the capability curve Perform G0-100-012, Power Operations for an unplanned power reduction Immediately reduce core power per the CRC instructions to lower power by 5 percent Perform G0-100-012, Power Operations for an unplanned power reduction Answer Explanation associated with the question as submitted ON-198-001 is the governing procedure for operation outside the generator capability curve with the main generator voltage regulator in AUTO. The initial conditions presented show operation just inside the limits of the capability curve. The transient results in sustained operation outside of the capability curve. A is incorrect.

While the AUTO voltage regulator has automatic circuitry to lower field current< 5876 amps, this is only activated on a generator field overvoltage condition, which has not occurred.

Adjusting the manual voltage regulator to match the AUTO regulator can be performed, but will not mitigate operation outside of the capability curve. B is incorrect.

Placing the manual voltage regulator in MANUAL is not authorized by the procedure.

There is no basis for assuming misoperation of the voltage regulator in AUTO as the stem clearly indicates the excessive reactive loading is due to grid conditions.

C is correct. A power reduction is authorized by ON-198-001.

Performing the power reduction per the CRC instructions is the preferred method. G0-1 00-012 will have to be performed due to the unplanned power reduction.

D is incorrect.

While a power reduction is authorized by ON-198-001, 5 percent is more than required to obtain a positive margin on the capability curve. Note 2 to Step 3.5.3 of ON-198-001 allows up to 2 minutes for the AUTO voltage regulator to attempt to restore margin, so immediate action is not required.

The 5 percent requirement is taken from ON-193-001 for a EHC control valve oscillation.

Page 2 of 13 Examination Analysis SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS 7 of 11 applicants selected response Cas the correct answer. 1 applicant selected D as the correct answer. 2 applicants selected A and 1 selected B. Recommended Change Accept response D as a second correct answer to the question, in addition to the original keyed answer C. Justification For Change -refer to reference diagrams below ON-198-001, Unit 1 Main Generator MVAR Control for AUTO Voltage Regulator Operation when Synched to the Grid, requires the operator to maintain the Main Generator parameters within those established by the Generator Capability Curve. Section 3.5 is directed to be performed if MVAR output is approaching or exceeding the curve limits. The information provided in the stem of the question establishes conditions outside the curve limits. Oscillations between 200 to 300 MVAR places the generator operation above the 75 psig curve limit as indicated on the plant computer reference given to the applicants.

Correctly applying the procedure requirements of ON-198-001 step 3.5.3 and G0-1 00-001, step 5.3.11.b, results in reducing reactor power which reduces generator power to restore operation within the limits of the curve. These procedural steps however do not indicate the amount of adjustment necessary.

On-198-001 does direct restoring operation to the allowable region as described in step 3.5 NOTE (2) d. Selecting response C, Reduce core power per the CRC instructions to lower generator load to restore positive margin to the capability curve and Perform 00-012, Power Operations for an unplanned power reduction as the correct answer results in a reduction of generator MW to restore "positive margin." The magnitude of positive margin is not described by the procedure requirements.

Interpolation of the numbers on the curve given results in a reduction to approximately 1300 MW to reenter the allowable region. Selecting response D, Immediately reduce core power per the CRC instructions to lower power by 5 percent Perform G0-1 00-012, Power Operations for an unplanned power reduction as the correct answer results in a reduction of generator MW to restore "positive margin" also. Lowering core power by 5 percent gives a corresponding decrease of generator power by 5 percent resulting in a final output power of 1272 MW. The magnitude of positive margin is not described by the procedure requirements.

Interpolation of the numbers on the curve provided to the applicants results in a reduction to approximately 1272 MW providing acceptable margin which would be reasonable for the conditions given. Directing a power reduction of 5 percent would be an acceptable operating practice as the SRO would be expected to establish a target power level when ordering a reactivity change. Note: SSES does not agree with the original question explanation of choice D which stated that ON-198-001 allows up to 2 minutes to restore margin, so immediate action is not required.

ON-198-001, Note (2), step 3.5.3 states, "Allow the Automatic Voltage Regulator no greater than two (2) minutes to reduce MVAR output if MVAR output exceeds Generator Capability Curve." Therefore, immediate action to reduce reactor core power is required and would be expected.

Response D is a subset of response C as a 5 percent reduction provides an operating point within the allowable range. An evaluation of the bases for selecting either CorD as an acceptable response reveals that in both cases the generator is operating within the established limits of the capability curve and all procedure requirements are followed.

Note: Distractors A and B remain incorrect for the reasons provided previously in Explanation section of the original question.

Given an evaluation of the above criteria SSES recommends accepting both responses C and D as correct answers. Page 3 of 13

References:

SSES LOC-26 NRC EXAMINATION SSES POST-EXAMINATION COMMENTS ON-198-001, Unit 1 Main Generator MVAR Control for AUTO Voltage Regulator Operation when Synched to Grid Generator Capability Curve and G0-1 00-012, Power Maneuvers Page 4 of 13 t='*'"m**OefWPF!WWWJj

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2!.1 SRO QUESTION 78 SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS Unit 1 is shutting down for a forced outage. Reactor Power is 20 percent. Annunciator TURB GEN BRG HI VIBRATION (AR-105-E05) alarms due to bearing #5 rotor and casing high vibration.

Operators trip the Main Turbine. The generator output breaker opens , but turbine speed does not lower. Turbine bearing #5 vibration continues to rise. Vibration is currently 15 mils, up 5 mils every minute. Which one of the following identifies the appropriate actions to direct to lower turbine vibration?

A. Close the MSIVs and MSL drains immediately Verify turbine speed begins to lower B. Place the Mode switch to SHUTDOWN immediately Close the MSIVs and MSL drains Verify turbine speed begins to lower C. Place the Mode switch to SHUTDOWN immediately Close the MSIVs and MSL drains Open the Main Condenser vacuum breakers if vibration does not lower D. Place the Mode switch to SHUTDOWN immediately Close the MSIVs and MSL drains Open the Main Condenser vacuum breakers after turbine speed lowers below 1200 rpm Answer Explanation associated with the question as submitted The Main Turbine has been tripped due to a high vibration condition.

On the turbine trip leak by on the main turbine stop and control valves has resulted in the turbine remaining at speed. Turbine vibration remains high and is rising slowly. A is incorrect.

Action to isolate steam flow to the main turbine is required by ON-193-002 Step 3.2. Although reactor power is below the bypass for reactor scram on turbine trip, directing an action, closing MSIVs that will result in a reactor scram without first initiating a reactor scram is not allowed. B is incorrect.

Per ON-193-002 Step 3.2 the steam supply to the main turbine should be isolated if turbine speed does not lower after a turbine trip. Additional action to break vacuum is warranted at this time due to the rapid rise in vibration.

C is correct. Breaking vacuum is required because vibration is rising rapidly. Vibration is above the trip limit , so the " extremely high" threshold has been met. This is the procedural method for breaking vacuum per ON-193-002. D is incorrect.

While the actions specified are correct and in the correct sequence, the trend on turbine vibration will result in extremely high values before the turbine coasts down to 1200 rpm. Page 8 of 13 Examination Analysis SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS 9 of 11 applicants selected response D as the correct answer. Recommended Change Accept distractor D as a second correct answer to the question, in addition to the original keyed answer C. Justification For Change ON-193-002, Main Turbine Trip, is entered as a result of the conditions given in the stem of the question.

The stem indicates that turbine speed is not lowering after the operators have tripped the turbine. Also turbine vibration continues to rise. The question asks the applicant to identify the appropriate actions to direct to lower turbine vibration

.. Correctly applying the procedure requirements of ON-193-002 step 3.4 results in scramming the reactor and closing the MSIVs and Main Steam Line Drains to lower turbine speed by isolating all steam flow to the turbine. The step also reduces the coast down time by increasing condenser pressure as a result of opening the main condenser vacuum breakers.

A Caution precedes step 3.4 indicating that "breaking vacuum when turbine speed is greater than 1200 rpm will cause severe stress on last stage turbine buckets." Selecting response Cas a correct choice accomplishes the following:

Place the Mode switch to SHUTDOWN immediately Close the MSIVs and MSL drains Open the Main Condenser vacuum breakers if vibration does not lower Based on information given in the stem of the question-Annunciator TURB GEN BRG HI VIBRATION (AR-1 05-E05) alarms due to bearing #5 rotor and casing high vibration, the applicant could correctly conclude from the alarm response procedure that an automatic turbine trip should have occurred at 11.5 mils but based on the stem conditions the turbine is still operating at normal speed. Breaking vacuum is required because vibration is rising rapidly as indicated in the stem of the question (currently 15 mils, up 5 mils every minute). Vibration is above the trip limit, so the "extremely high" threshold as stated in ON-193-002, step 3.4 has been met. Therefore, selecting step 3.4 of ON-193-002 is a correct response to the information given in the stem of the question. It is reasonable for the applicant to correctly conclude that isolating steam to the turbine will reduce turbine speed and vibration.

The last element of response C qualifies the action to be taken to "Open the Main Condenser vacuum breakers only if vibration does not lower". This qualifying statement about the effect of the two previous actions (i.e., if vibration does not lower) could lead the applicant to direct opening the main condenser vacuum breakers without applying the caution statement (i.e., to delay until turbine speed lowers to 1200 rpm) since reduced coast down time is required if vibration does not lower. This qualifying statement (i.e., if vibration does not lower) makes response C correct by supplying the applicant with the conditions of the turbine after the first two actions are taken. Page 9 of 13 SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS Selecting response D as a correct choice accomplishes the following:

Place the Mode switch to SHUTDOWN immediately Close the MSIVs and MSL drains Open the Main Condenser vacuum breakers after turbine speed lowers below 1200 rpm. The first two actions in answer choices C and D are correct actions from ON-193-002 step 3.4. The decision to apply the caution is based on whether turbine vibrations will lower after closing the MSIVs. It is operationally valid for the applicants to conclude that vibrations will lower following closure of the MSIVs. The applicant could correctly conclude that isolating steam to the turbine will reduce turbine speed with a resultant reduction in vibration.

Unlike answer choice C, no qualifying information is given in answer choice 0 regarding the status of turbine vibration following closure of the MSIVs. Therefore, the operator would be correct in applying the procedural caution statement to delay opening the vacuum breakers until turbine speed lowers to 1200 rpm. Since no qualifying information is given regarding turbine vibration following MSIV closure in response 0, response D is also a correct response.

Note: Answer choices A & B are incorrect.

A is incorrect because if the mode switch was not taken to shutdown prior to closing the MSIVs an automatic high pressure scram would occur. B is incorrect because the stem of the question asks which of the following identifies the appropriate actions to direct to lower turbine vibration.

Answer choice B states "Verify turbine speed begins to lower. In selecting response B the applicant must disregard the information given in the stem which indicates turbine vibration is extremely high and rising (i.e., so high that the turbine should have tripped).

ON-193-002, step 3.4, requires reduced coastdown time since a condition of extremely high vibrations exits requiring breaking vacuum. Additional action is required which would result in breaking condenser vacuum either immediately or when turbine speed lowers to 1200 rpm. It's incumbent upon the operator in this situation to reduce the speed of the turbine in a timely fashion thus reducing turbine vibration as quickly as possible.

Selecting choice B does not reduce speed in a timely manner which is directed by ON-193-002.

Given an evaluation of the above criteria SSES recommends accepting responses C and D both as correct answers.

References:

ON-193-002, Main Turbine Trip AR-1 05-001 ,EOS Page 10 of 13 SRO QUESTION 80 Question SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS Unit 1 has experienced a failure of RPS to trip. When ARI was initiated, a large number of control rods on the right side of the full core display continued to show not fully inserted All actions in the power leg of E0-1 00-113 were completed to the point of attempting control rod insertion.

ES-158-002, ARI and RPS Trip Bypass, was directed to be performed.

The in-field portion of the ES was completed. Annunciators RPS CHAN A1/A2(B1/B2)

SCRAM DSCH VOL HI WTR LEVEL TRIP (AR-103(104)-F02), have subsequently cleared. Which one of the following should be directed next in an attempt to insert the withdrawn rods? A. Reset the scram, then insert a manual scram using the RPS manual scram pushbuttons in accordance with ES-158-002, RPS and ARI Trip Bypass B. Individually scram control rods in accordance with Attachment A of E0-1 00-113 Sheet 2 C. Vent the scram air header in accordance with the posted instructions D. Insert control rods in accordance with ES-155-001, Venting CRD to Insert Control Rods Answer Explanation associated with the question as submitted The conditions presented in the stem are consistent with an electrical A TWS, as indicated by the failure of the full core display to enter full-in/full-out mode, where ARI initiation or maximizing CRD flow were successful in inserting most of the control rods. ES-158-002 was directed for installation to defeat ARI to re-pressurize the scram air header for subsequent scram attempts.

The RPS trip bypass portion of the ES were installed, but for no effect. A is incorrect.

The actions described are the next steps to perform to complete ES-158-002 to attempt a re-scram.

However, as RPS has failed to trip this action will not have any effect and will not insert control rods. B is incorrect.

Individually attempting to scram control rods may have some success, but re-venting the scram air header to attempt to re-scram all withdrawn control rods is the preferred response.

C is correct. With RPS untripped and ARI defeated to re-pressurize the scram air header (indicated by the SDV now being drained), this action will vent the scram air header for an attempt to re-scram the control rods. D is incorrect.

This action would be effective in attempting to insert the control rods, but is not allowed to be used until all other methods have been attempted.

Page 11 of 13 Examination Analysis SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS 10 of 11 applicants missed the question overall. 6 applicants selected response B, 4 applicants selected response A. Recommended Change Accept distractor Bas a second correct answer to the question, in addition to the original keyed answer C. Justification For Change In order to select the correct response, the applicant must determine the malfunctions in the automatic scram system and align the remaining available methods of control rod insertion with those not affected by the malfunctions described in the stem of the question. The applicant could correctly conclude the following from the information given in the stem:

• RPS has completely failed (stated in stem)

• ARI was partially effective (stated in stem that a large number of control rods did not show fully inserted)

-based on this information the ARI valves actuated, the scram air header depressurized, and some control rod movement occurred but was hindered by some unstated hydraulic interference.

• Emergency Operating Procedures were entered (E0-1 00-113 SH2, Control Rod Insertion) and steps completed up to the point of attempting control rod insertion (stated in the stem)

• ES-158-002, ARI and RPS Trip Bypass in field portion completed.

Based on the action accomplished by ES-158-002, ARI is bypassed and will no longer actuate from the control room switches.

Note the procedure would also bypass RPS except for manual scram however; this information is not relevant since the information in the stem establishes a failure of RPS to trip under any circumstance.

• The scram condition originally inserted by ARI (which was partially effective) has been reset and the scram discharge volume is drained based on RPS CHAN A 1/A2 (81/82) SCRAM OSCH VOL HI WTR LEVEL TRIP being clear which is stated in the stem. Selecting response C (Vent the scram air header in accordance with the posted instructions), as a correct choice is based on correct conclusions made by the applicant as follows: This method of control rod insertion is an approved method directed by procedure (E0-1 00-113 SH2, Control Rod Insertion).

Note that the procedural paths in E0-1 00-113 SH2 are not prioritized.

Methods are valid if operationally possible based on the system malfunction present.

• The method of venting the scram air header is an operationally valid choice since ARI and RPS cannot be used from the control room to initiate a scram, and since the scram air header is re-pressurized as concluded from the scram discharge volume vent and drain valves being open and the scram discharge volume being drained, therefore this method to vent the scram air header locally to insert control rods could be effective.

• This method of venting the scram air header to insert control rods was partially effective before since ARI had previously been initiated and some control rod movement occurred as indicated by information provided in the stem.

• The physical plant locations to vent the scram air header (choice C) or individually scram the control rods (choice B) are the same (locally near the hydraulic control unit banks in the reactor building) therefore both of these actions could be completed in parallel if sufficient manpower was available.

Based on this evaluation, the applicant could conclude that response Cis a correct choice. Page 12 of 13 SSES LOC-26 NRC EXAMINATION SSES POST -EXAMINATION COMMENTS Selecting response B (Individually scram control rods in accordance with Attachment A of E0-1 00-113 Sheet 2), is also a correct choice based on correct conclusions made by the applicant as follows:

• The method of control rod insertion is an approved method directed by procedure (E0-1 00-113 SH2, Control Rod Insertion) and all conditions given in step CR-16 are met to proceed as described below and indicated by the information given in the stem. IF ALL

• ARI DISABLED

• SCRAM RESET

• SCRAM AIR HEADER NOT VENTED

• RPS FUSES NOT REMOVED Note that the procedural paths in E0-1 00-113 SH2 are not prioritized.

Methods are valid if operationally possible based on the system malfunction present. The method of individually scramming control rods is an operationally valid choice since ARI and RPS cannot be used from the control room to initiate a scram, therefore another method to individually scram control rods is appropriate.

• The applicant concludes that the method of venting the scram air header to insert control rods was only partially effective before since ARI had previously been initiated and a large number of control rods did not indicate fully inserted.

• The physical locations to individually scram the control rods (choice B) or vent the scram air header (choice C) are the same (locally near the hydraulic control unit banks in the reactor building) therefore personnel travel time to complete the actions or personnel safety would not be a factor in selecting an alternate method.

• The bases for E0-100-113 SH2, Control Rod Insertion, document E0-000-113 step CR 16 for individually scramming control rods states "Even with an incompletely drained or non-vented scram discharge volume, sufficient capacity may remain in the scram discharge volume to accommodate discharge flow from control rod scrams taken one at a time as opposed to all at once." The applicant could correctly conclude that this method would be effective based on the information provided in the stem of the question.

• The bases for E0-1 00-113 SH2, Control Rod Insertion, document E0-000-113 step CR 16 for individually scramming control rods also states Control rods are individually scrammed until control rods will not insert and it is determined that other methods of inserting control rods would be more effective.

Based on the information given in the stem, there is no remaining information for the applicant to determine if another method would be more effective until this method is applied since the initial indication given was that the first method of scramming was only partially effective.

Based on this evaluation, the applicant could conclude that response B is a correct choice. Responses A & D remain incorrect.

Response A cannot be a correct response because the stem states RPS failed and therefore cannot be tripped to initiate control rod insertion.

Response D cannot be a correct response because other methods are available (choices B & C), and the bases for E0-100-113 SH2, Control Rod Insertion, document E0-000-113 step CR 28 states that this method for venting CRD over piston volumes is only applicable when "ALL OTHER METHODS HAVE FAILED" and further explains that this step is entered when all other means of inserting control rods have been unsuccessful.

Given an evaluation of the above criteria SSES recommends accepting responses B and C both as correct answers. References E0-1 00-113 SH2, Control Rod Insertion and EOP Bases for steps CR-16 and CR-28 ES-158-002, ARI and RPS Trip Bypass Page 13 of 13 PROCEDURE COVER SHEET PPL SUSQUEHANNA, LLC PROCEDURE 6/13/2014 MAIN TURBINE TRIP ON-193-002 Revision 24 I Page 1 of 1444 ADHERENCE LEVEL: CONTINUOUS USE QUALITY CLASSIFICATION:

APPROVAL CLASSIFICATION: (X) QA Program ( ) Non-QA Program (X) Plant ( ) Non-Plant ( ) Instruction EFFECTIVE DATE: PERIODIC REVIEW FREQUENCY:

2 Year PERIODIC REVIEW DUE DATE: RECOMMENDED REVIEWS: Procedure Owner: F Shift Responsible Supervisor:

Shift Manager-F Shift Responsible FUM: Manager-Nuclear Operations Responsible Approver: Manager-Nuclear Operations

1. SYMPTOMS AND OBSERVATIONS ON-193-002 Revision 24 Page 2 of 14 1.1 Any of following alarms at Unit Operating Bench board 1 C651: 1.1.1 MAIN TURB MASTER TRIP 1.1.2 GEN LOCKOUT RELAYS TRIP 1.1.3 EXCITER FIELD BKR TRIP 1.2 One or more annunciators associated with condition causing Main Turbine/Main Generator trip at Unit Operating Benchboard 1 C651. 1.3 Reactor Scram if Main Turbine trip occurred with Reactor power> 26%. 2. AUTOMATIC ACTIONS 2.1 Main Turbine Stop Valves SV-1 ,2,3 and 4 close. 2.2 Main Turbine Control Valves CV-1 ,2,3 and 4 close. 2.3 Main Turbine Combined Intermediate Valves CIV-1 ,2,3,4,5 and 6 close. 2.4 Bypass Valves BPV-1, 2, 3, 4 and 5 open to maintain set pressure. 2.5 Auxiliary Busses 11A (1A 101) and 11 B (1A 1 02) transfer to Startup Bus 10 (OA 1 03). 2.6 As Main Turbine coasts down, following pumps start: 2.6.1 Motor Suction Pump 1 P1 08. 2.6.2 Turning Gear Oil Pump 1 P111. 2.6.3 Turbine Lift Pumps 1 P1 09A,B,C,D,E,F,G,H

& J 2.7 Reactor Recirc Pumps 1 P401A and B trip caused by EOC-RPT if thermal power >26%. 2.8 Stator Cooling Pumps 1 P116A and B trip if Main Turbine Tripped on Main Generator Lockout due to Generator Differential or Generator Neutral Overvoltage conditions (86GE Lockout Relay).

ON-193-002 Revision 24 Page 3 of 14 3. OPERA TOR ACTIONS D D D D D D D D D D D D D D D D NOTE: 3.1 Ensure: 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1.7 3.1.8 3.1.9 Subsections within section 3 may be performed in any order as determined by Shift Supervision based on the nature of the event and the priority of required operator actions. Steps within each subsection must be performed in the order written. Main Generator Output Breaker TRIPS. IF Main Generator is unable to disconnect from Grid after a Turbine trip, Perform ON-198-004.

Main Generator Exciter Field Breaker TRIPS. Auxiliary Bus "11A" 1A101 TRANSFERS to Startup Bus OA103. Auxiliary Bus "11 B" 1 A 102 TRANSFERS to Startup Bus OA 103. Motor Suction Pump 1 P108 STARTS. Turning Gear Oil Pump 1 P111 STARTS. Turbine Lift Pumps 1P109A,B,C,D,E,F,G, H & J START. IF Rx Power> 26% when Main Turbine Tripped, Reactor Recirc Pumps 1 P401A and B TRIP caused by EOC-RPT. 3.2 IF Main Turbine speed not lowering, Perform the following:

3.2.1 IF Rx Power< 26% when Main Turbine tripped, Scram the Reactor PER ON-100-101, Scram Scram Imminent.

3.2.2 Close

the MSIV's and Main Stm Drains

• MN STM LINE A OB ISO HV-141-F028A

• MN STM LINE BOB ISO HV-141-F028B

• MN STM LINE COB ISO HV-141-F028C

• MN STM LINED OB ISO HV-141-F028D

• MN STM LINE A IB ISO HV-141-F022A D D D D D D D 3.3 * * * *

• ON-193-002 Revision 24 Page 4 of 14 MN STM LINE BIB ISO HV-141-F022B MN STM LINE C IB ISO HV-141-F022C MN STM LINED IB ISO HV-141 F022D MN STM LINE IB DRAIN HV-141-F016 MN STM LINE OB DRAIN HV-141-F019 Monitor Main Turbine coastdown.

CAUTION Breaking vacuum when turbine speed is greater than 1200 rpm will cause severe stress on last stage turbine buckets. D D D D D D D D D 3.4 IF reduced coastdown time is required due to extremely high bearing vibration, Break vacuum as follows: NOTE: 3.4.1 3.4.2 MSIV's will close at 19.0 inches Hg Absolute (10.2 inches Hg Vacuum) unless Low Vacuum Trip bypassed.

Turbine Bypass Valves will close at 22.2 inches Absolute (7 inches Hg Vacuum). IF Rx Power< 26% when Main Turbine tripped, Scram the Reactor PER ON-1 00-101 , Scram Scram Imminent.

Close the MSIV's and Main Stm Drains

• MN STM LINE A OB ISO HV-141-F028A

• MN STM LINE BOB ISO HV-141-F028B

• MN STM LINE COB ISO HV-141-F028C

• MN STM LINED OB ISO HV-141-F028D

• MN STM LINE A IB ISO HV-141-F022A

• -MN STM LINE BIB ISO HV-141-F022B

• MN STM LINE C IB ISO HV-141-F022C D D D D D D D D D D D 3.4.3 3.4.4 3.4.5 ON-193-002 Revision 24 Page 5 of 14

• MN STM LINED IB ISO HV-141F022D

• MN STM LINE IB DRAIN HV-141-F016

• MN STM LINE OB DRAIN HV-141-F019 At 1C668 , Open CDSR VAC BKR HV-10742A, B , C by depressing common OPEN push button. Observe Main Condenser Vacuum decays to atmospheric pressure.

Remove Sealing Steam from turbine Seals in accordance with OP-192-00

1. 3.5 Evaluate need to enter following ON's: 3.5.1 ON-100-101, Scram , Scram Imminent 3.5.2 ON-164-002 , Loss of Reactor Recirculation Flow 3.5.3 ON-197 -001, Loss of Stator Cooling 3.6 At Generator and Transformers Protection Relays Vertical Board 1 C654, Circle as found position AND as applicable Reset the following Primary and Backup Lockout Relays if tripped: NOTE: Actuation of Aux Transformer 11 Ground Overvoltage Lockout Relay (86GE) indicates a possible ground fault condition on either the 11A or 11 B Aux Buses. 86GE Relay actuation blocks the Fast and Slow transfer of Aux Buses to Tie Bus. CAUTION Resetting the Aux Transformer 11 Neutral Overvoltage Relay (86GE) could re-energize the Fault condition.

D D 3.6.1 IF Aux Transformer 11 Neutral Overvoltage Relay 86GE Tripped: a. b. Contact Electrical Maintenance for investigation.

DO NOT Reset Auxiliary Transformer 11 Neutral Overvoltage Lockout Relay 86GE.

D D D D D D D D D D NOTE: c. ON-193-002 Revision 24 Page 6 of 14 DO NOT Reset any Generator Lockout Relays, until after the cause of the 86GE Lockout Relay trip is understood.

Primary Lockout Relays 86GA and 86GC cannot be reset until Gen Sync BKR Low Gas Pressure Lockout condition is cleared at 1R101. (AR-106-E10) 3.6.2 IF Aux Transformer 11 Neutral Overvoltage Relay 86GE Reset: a. Reset Generator Primary Lockout Relays: ( 1) (2) 86GA TRIPPED/RESET (circle as found position) 86GC TRIPPED/RESET (circle as found position) 3.6.3 IF Backup Lockout Relays 86GB and 86GD Tripped, a. Reset/Ensure Reset Breaker Failure Lockout Relay 86BF-190B1 at Panel OC190B (A13, El714). NOTE: Backup Lockout Relays 86GB and 86GD cannot reset until Breaker Failure Lockout Relay 86BF-190B 1 at Panel OC190B is reset. b. Reset Generator Backup Lockout Relays: ( 1) (2) 86GB TRIPPED/RESET (circle as found position) 86GD TRIPPED/RESET (circle as found position)

3. 7 At Generator and Transformer Protection Relays Vertical Board 1 C654, Perform the following for Stator Coolant Pumps 1 P116A(B).

3. 7.1 Circle as found position for following Primary and Backup Stator Cooling Lockout Relays. a. 86GI TRIPPED/RESET

b. 86GJ TRIPPED/RESET D D ON-193-002 Revision 24 Page 7 of 14 3.8 Perform one of the following: 3.8.1 3.8.2 IF turbine tripped due to a rub during turbine startup, WHEN directed by Shift Supervision , Return turbine to 100 RPM lAW OP-193-001, MAIN TURBINE STARTUP AND Continue with this procedure.

At ZERO RPM, Ensure Turning Gear 1S103 AUTOMATICALLY STARTS and ENGAGES at TURNING GEAR HS-10168.

D 3.9 IF Main Turbine will be out of service for> 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, Remove Turbine from Turning Gear lAW OP-193-001.

D 3.10 WHEN Plant conditions stabilize, Refer to General Operating Procedure for Plant D D cooldown or startup to place Plant in desired conditions.

3.11 At Turbine EHC Panel 1 C663 (Bay E) 3.11.1 3.11.2 PRIOR to Reset, Record First Hit: __ _ Reset First Hit by Depressing the two reset pushbuttons simultaneously.

D 3.12 Circle relay target numbers operated on Attachment A , Relay Target Status Sheet. D NOTE: Section 5.2 describes relay functions.

D 3.13 Reset tripped relay targets. D 3.14 Ensure Main Generator shutdown in accordance with OP-198-001 , Main Generator System. D 3.15 IF Tripped, Reset following Primary and Backup Stator Cooling Lockout Relays D AFTER determination made that Stator Cooling System operation will not adversely impact Main Generator performance.

3.15.1 Ensure selector switch HSS-10186 at Generator Hydrogen and Stator Cooling Water Control Panei1C125 to 1A-1B/STOP (1 B-1A/STOP) position.

D D D D NOTE: 3.15.2 3.15.3 ON-193-002 Revision 24 Page 8 of 14 Stator Coolant Pumps 1 P116A and 1 P116B will auto start in the next step unless selector switch HSS-1 0186 is positioned for Stator Cooling Pump Shutdown. Reset Primary and Backup Stator Cooling Lockout Relays: a. 86GI RESET b. 86GJ RESET IF required, Return Stator Cooling System to service in accordance with OP-197 -001. 3.16 Forward completed copy of this procedure to following for review: 3.16.1 Unit Supervisor I Signature Date 3.16.2 Shift Manager I Signature Date 3.16.3 System Engineering

-Switchyards I Signature Date 3.16.4 Assistant Operations Manager 3.17 Forward reviewed copy of this procedure to: None 4. REFERENCES

4.1 Electrica'l

Schematic E-14 Sh 1 Logic Diagram Unit 1 Protection

4.2 Electrical

Schematic E-21 Sh 1 Schematic Meter Relay Diagram Generator Main Unit Aux Trans Unit 1 4.3 Electrical Schematic E-111 Sh 1-4 Schematic Diagram Unit Protection Unit 1 4.4 FSAR Section 15.2.3 Turbine Trip 4.5 EWR 1549711

5. DISCUSSION ON-193-002 Revision 24 Page 9 of 14 This procedure addresses actions for any cause of Main Turbine Trip. For a Main Turbine trip above 26% Reactor power, based on Turbine First Stage Pressure, Stop Valve Closure or Control Valve fast closure causes Reactor Scram. Below this value, Scram is bypassed.

Main Turbine Trip can be caused by Main Generator Trip on Load Reject or other electrical fault. These trips protect Main Generator from damage. For Main Turbine Trip accompanied by Reactor Scram, actions of this procedure must be performed in conjunction with ON-1 00-101, Scram, Scram Imminent.

5.1 The following abnormal conditions or malfunctions produce a Main Turbine trip: 5.1.1 Unit Primary Lockout Relay Trip 5.1.2 Unit Backup Lockout Relay Trip 5.1.3 Low Electro-Hydraulic-Control (EHC) pressure 5.1.4 Thrust Bearing Wear Detector 5.1.5 Low Bearing Oil Pressure 5.1.6 Moisture Separator Drain Tank High Level 5.1.7 Turbine High Vibration 5.1.8 Loss of Stator Cooling 5.1.9 Shaft Lube Oil Pump Low Discharge Pressure 5.1.10 Loss of Primary and Backup Speed Signal 5.1.11 Low Condenser Vacuum 5.1.12 Overs peed 5.1.13 Backup Overspeed 5.1.14 High Reactor Vessel Level 5.1.15 Manual ON-193-002 Revision 24 Page 10 of 14 5.2 Following abnormal conditions or malfunctions produce Main Generator Primary and Backup Lockout causing Main Turbine trip. PANEL 1C654 DESCRIPTION TRIPS 86G TARGET NOs LOCKOUTS 1 & 2/2 MAIN TRF 1A 0A DIFF A,C 2 & 2/2 MAIN TRF 1A 0A INSTANT A,C 3 & 2/2 MAIN TRF 1A 0B DIFF A , C 4 & 2/2 MAIN TRF 1A 0B INSTANT A , C 5 & 2/2 MAIN TRF 1A 0C DIFF A , C 6 & 2/2 MAIN TRF 1A 0C INSTANT A,C 7 &2/3 MAIN TRF 1 B 0A DIFF A,C 8 & 2/3 MAIN TRF 1B 0A INSTANT A,C 9 & 2/3 MAIN TRF 1 B 0B DIFF A,C 10 & 2/3 MAIN TRF 1B 0B INSTANT A,C 11 & 2/3 MAIN TRF 1B 0C DIFF A,C 12 & 2/3 MAIN TRF 1B 0C INSTANT A,C 13 & 3/3 AUX TRF 11 0A DIFF A , C 14 & 3/3 AUX TRF 11 0A INSTANT A , C 15 & 3/3 AUX TRF 11 0B DIFF A , C 16 & 3/3 AUX TRF 11 0B INSTANT A,C 17 & 3/3 AUX TRF 11 0C DIFF A,C 18 & 3/3 AUX TRF 11 0C INSTANT A,C 19 & 1/4 ALTERNATOR EXCITER A0 DIFF A , C 20 & 1/4 ALTERNATOR EXCITER B0 DIFF A , C 21 & 1/4 ALTERNATOR EXCITER C0 DIFF A,C 22 & 1/1 GENERATOR 0A DIFF A , C , I 23 & 1/1 GENERATOR 0B DIFF A,C,I 24 & 1/1 GENERATOR 0C DIFF A,C,I 25 & 3/5 GEN NEGATIVE PHASE SEQUENCE A , C 26 and alarm GEN VOLTAGE UNBALANCE RELAY & GEN REGULATOR POTENTIAL FUSE BLOWN NONE AR106 H06 VOLT/REG/PRIMARY RELAY POTENTIAL FUSE BLOWN 27 and alarm GEN VOLTAGE UNBALANCE RELAY & GEN METERING FUSE BLOWN NONE AR106 G06 METER/BACKUP RELAY POTENTIAL FUSE BLOWN 28 & 3/4 TRF 11 GROUND DIFF A , C 29 & 2/4 GEN PRIMARY ANTIMOTORING A , C ON-193-002 Revision 24 Page 11 of 14 PANEL 1C654 DESCRIPTION TRIPS 86G TARGET NOs LOCKOUTS 30 & 1/3 GEN LOSS OF FIELD A,C 31' 32, 33, & 1/5 GEN UNDER FREQUENCY A,C 34 & 1/2 GEN NEUTRAL OVERVOL TAGE A,C,I 35 & 2/1 230 KV SPAN 0A DIFF A,C 36 & 2/1 230 KV SPAN 0B DIFF A,C 37 & 2/1 230 KV SPAN 0C DIFF A,C 38 & 5/1 UNIT 0A DIFF B,D 39 & 5/1 UNIT 0A INSTANT B,D 40 & 5/1 UNIT 0B DIFF B,D 41 & 5/1 UNIT 0B INSTANT B,D 42 & 5/1 UNIT0CDIFF B,D 43 & 5/1 UNIT 0C INSTANT B,D 44 & 4/1 GEN OUT OF STEP B,D 45 & 5/3 AUX TRF 11 0A OVERCURRENT B,D 46 & 5/3 AUX TRF 11 0A INSTANT B,D 47 & 5/3 AUX TRF 11 0B OVERCURRENT B,D 48 & 5/3 AUX TRF 11 0B INSTANT B,D 49 & 5/3 AUX TRF 11 0C OVERCURRENT B,D 50 & 5/3 AUX TRF 11 0C INSTANT B,D 51 & 4/2 GEN GROUND OVERVOL TAGE B,D,J 52 & 4/3 GEN NEUT OVERVOLTAGE (STARTUP)

B,D,J 53 & 5/2 AUX TRF 11 NEUT OVERVOL TAGE B,D,E 54 and alarm GEN VOLTAGE UNBALANCE RELAY & GEN METERING POTENTIAL FUSE BLOWN NONE AR106 G06 METER/BACKUP RELAY POTENTIAL FUSE BLOWN 55 and alarm GEN VOLTAGE UNBALANCE RELAY & GEN REGULATOR POTENTIAL FUSE BLOWN NONE AR106 H06 VOLT/REG/PRIMARY RELAY POTENTIAL FUSE BLOWN 56 & 5/4 GEN LOSS OF FIELD B,D 57 & 6/1 GEN BACKUP ANTI-MONITORING B,D 2/5 GEN OVEREXCITATION (110%) A,C 3/1 MAIN TRF 1A SUDDEN PRESSURE A,C 3/2 MAIN TRF 1B SUDDEN PRESSURE A,C 4/4 GEN OVEREXCITATION (118%) B,D PANEL 1C654 TARGET NOs 4/5 5/5 6/2 6/3 6/4 6/5 DESCRIPTION 230 KV SPAN PROTECTION ON-193-002 Revision 24 Page 12 of 14 TRIPS 86G LOCKOUTS B,D (SEE TARGETS ON OC190A, OC190B , OC192) SPARE 230 KV SPAN PROTECTION A,C (SEE TARGETS ON OC190A, OC190B , OC192) GENERA TOR OVERSPEED A,C BREAKER FAILURE OR OPEN BREAKER B , D FLASHOVER (SEE TARGETS ON OC190B) GENSYNCBKRLOSSGASLOCKOUT TARGET NOs Panel OC190A 1 Line Diff 87-190A01 2 Prim Trip Timer 62P-190A01 DESCRIPTION 3 POTT 21B/67NB-190B01 Panel OC190B 4 B.U. Trip Timer 62B-190B01 Panel OC192 5 Breaker Failure 50-190B01 6 Breaker Failure 86BF-190B01 21 CARRIER 2 START ON-193-002 Revision 24 Page 13 of 14 TRIP 86G LOCKOUTS A, B, C, D A, B, C, D A, B, C, D A, B, C, D A,B,C,D A, B, C, D 8 DTT1 Trip Receive 9 TCO Trip Receive 67N CARRIER 2 GROUND TRIP & START A,B,C,D A,B,C,D 5.3 Backup Lockout Relay 5.3.1 Generator Out of Step 5.3.2 Power Directional Relay for Anti-motoring 5.3.3 Loss of Field Relay Backup 5.3.4 Unit Auxiliary Transformer Overcurrent 5.3.5 Unit Auxiliary Transformer Neutral Overvoltage

5.3.6 Generator

Neutral Overvoltage 5.3.7 Unit Differential

5.3.8 Generator

Ground Overvoltage

5.3.9 Generator

Overexcitation 5.3.10 Generator 230KV Circuit Breaker Failure Attachment A ON-193-002 Revision 24 Page 14 of 14 UNIT 1 RELAY TARGET STATUS SHEET PANEL 1C654 1 11 21 31 41 51 111 211 311 411 511 611 2 12 22 32 42 52 112 212 312 412 512 612 3 13 23 33 43 53 113 213 313 413 513 613 4 14 24 34 44 54 114 214 314 414 514 614 5 15 25 35 45 55 115 215 315 415 515 615 6 16 26 36 46 56 7 17 27 37 47 57 86GA 86GB 8 18 28 38 48 86GC 86GD 9 19 29 39 49 86GI 86GE 10 20 30 40 50 86GJ I I PCO Time Date Panel OC190A 1 2 Panel OC190B 3 4 5 6 Panel OC192 8 9 I I NPO Time Date Page 1 of 1 PROCEDURE COVER SHEET PPL SUSQUEHANNA, LLC PROCEDURE 3/13/2014 LEVEL/POWER CONTROL E0-000-113 Revision 13 Page 1 of 67 ADHERENCE LEVEL: CONTINUOUS USE QUALITY CLASSIFICATION:

APPROVAL CLASSIFICATION: (X) QA Program ( ) Non-QA Program (X) Plant ( ) Non-Plant ( ) Instruction EFFECTIVE DATE: PERIODIC REVIEW FREQUENCY:

2 Year PERIODIC REVIEW DUE DATE: RECOMMENDED REVIEWS: Procedure Owner: EOP Coordinator Responsible Supervisor:

Supervisor-Operations Engineering Responsible FUM: Manager-Nuclear Operations Responsible Approver: Manager-Nuclear Operations

1. GENERAL E0-000-113 Revision 13 Page 2 of 67 Provides instructions to control reactor power, RPV water level , and RPV pressure during ATWS events. These three parameters are controlled in a manner which minimize the threat to both core and primary containment integrity. Core integrity is threatened by operation with low RPV water level (near or below TAF), low RPV pressure, and high core inlet subcooling.

Primary containment integrity is threatened by excessive thermal loading. Operation with water level near or below TAF is avoided because in this region level is expected to be difficult to control, there is potential for fuel uncovery and heat-up of the cladding , and only Fuel Zone level indication, which is not calibrated for reactor operation at rated pressure, is available. At low RPV pressure the reactor is expected to be highly unstable, and power oscillations may lead to fuel damage. Unstable operation with potential for fuel damage can also occur at high levels of core-inlet subcooling even with the reactor at rated pressure.

In the non-isolation ATWS (MSIVs open), containment integrity is threatened by fission power in excess of the turbine bypass system capacity. For the isolation ATWS , fission power in excess of the RHR system capacity presents a threat to containment.

Actions which are designed to limit threat to core and containment include:

• Optional methods of inserting control rods and injection of boron.

• Reducing reactor power and core-inlet subcooling by deliberately lowering RPV water level.

• Assigning a priority to the use of injection systems.

• Delaying RPV cooldown until reactor shutdown has been assured by control rod insertion or injection of Cold Shutdown Boron Weight.

• Avoiding reactor depressurization on PSL and HCTL if initial ATWS power was greater than 5%.

• Raising RPV water level to remix boron. Generic tracking tables for ES procedures and parameters are on the board to assist the operating crew in track i ng parameter trends and ES procedure implementation.

These were added to the board as aids, and do not trigger or direct activities.

2. ENTRY CONDITIONS The entry condition to this flowchart is: Any entry condition to E0-000-1 02 More than 1 control rod is > 00 E0-000-113 Revision 13 Page 3 of 67 If the scram is unsuccessful upon entry to E0-000-1 02, a failure-to-scram event (ATWS) has occurred and E0-000-102 step RC-2 directs entry to this flowchart.

TABLE 2 RPV CONTROL ENTRY CONDITIONS

• RPV WATER LVL < +13"

• RPV PRESSURE>

1087 PSIG

• DW PRESSURE>

1.72 PSIG

• EXISTING SCRAM CONDITION AND PWR > 5% OR CANNOT BE DETERMINED E0-000-1 02 may have been entered due to any one of the four plant parameter entry conditions listed in Table 2, or any one of the procedural entry conditions from Primary Containment Control, Secondary Containment Control, or Radioactivity Release Control. If while performing E0-000-113 actions, another RPV Control entry condition occurs, it is only necessary to re-enter E0-000-113 at Step LQ-1. During an ATWS event, re-entry to E0-000-102 is not required.

If direction to purposely lower level is in progress, re-entry on low RPV water level is not required.

Since lowering level is procedurally controlled, this does not represent a change in conditions that needs to be evaluated.

E0-000-113 effects control of RPV water level under conditions when it cannot be determined that control rod insertion alone will assure that the reactor will remain shutdown under all conditions.

Actions to control RPV water level in this flowchart differ from those in E0-000-1 02 in order to address four basic concerns:

1. When boron is injected into the RPV, the systems used for control of RPV water level must be operated to minimize boron dilution and cold water injection, and to promote boron mixing.

E0-000-113 Revision 13 Page 4 of 67 2. If the reactor cannot be shut down and suppression pool temperature continues to rise, RPV water level must be controlled not only to cool the core, but also to minimize suppression pool heatup. 3. In the non-isolation ATWS where the feedwater system is available for makeup to the RPV, water level must be controlled to prevent the buildup of high levels of core-inlet subcooling which can drive the reactor into unstable operation.

4. Even if boron has not been injected into the RPV and the reactor is subcritical on control rods under hot conditions, injection of cold water could cause criticality without negative reactivity

_ feedback occurring until reactor power reaches the heating range. (

Reference:

PSTG RPV Control Guideline Purpose and Entry Conditions)

3. PROCEDURE D NOTE: Subsections within Section 3 may be performed in any order as determined by Shift Supervision based on the nature of the event and priority of the operator actions. Steps within each subsection must be completed in the order written. LQ-1 EXIT RPV CONTROL When directed to control RPV parameters during ATWS conditions, guidance given in E0-000-113 supersedes instructions given in E0-000-1 02. It is therefore necessary to exit this procedure. (

Reference:

PSTG 1st Override before RC/L-2) LQ-2 IS MORE THAN 1 CONTROL ROD > 00 This step is applicable to all subsequent steps within this flowpath.

It is applicable whether the step was entered from above, or transfer resulted in entry below this step. It remains applicable to those steps until flowchart is exited. The intent of this step is to define the condition that allows E0-000-113 to be exited. E0-000-113 may only be exited if it can be determined the reactor will remain shutdown under all conditions without boron. If reactor shutdown margin can be assured for all conditions by control rod insertion alone, reactor power, RPV level and RPV pressure control actions are continued in E0-000-1 02.

E0-000-113 Revision 13 Page 5 of 67 The most effective way for an Operator to make this determination whether the reactor will remain shutdown under all conditions without boron is by observing that no more than one control rod is greater than position 00. If more than one control rod fails to full y insert, the operator can determine that the reactor will remain shutdown by confirming all control rods are inserted to the MSBWP (Maximum Subcritical Banked Withdrawal Position). MSBWP is the maximum banked control rod position to which all control rods can be withdrawn in a.ddition to the highest worth control rod being fully withdrawn and the reactor remain shut down under all condit i ons. MSBWP is 00 for Un it 1 and 00 for Unit 2. It is calculated assuming:

1. Most reactive time i n core life. 2. No xenon is present in the core. 3. No voids are present in the core. 4. Moderator temperature is 68°F. 5. Highest worth control rod is fully withdrawn.

If all control rods are inserted to the MSBWP , (in addition to the highest worth control rod being fully withdrawn), the reactor will remain shutdown and the intent of this step is met. All other EO steps that also confirm the reactor is shutdown can be answered " NO." For control rod configurations where more than one control rod is at a position greater than MSBWP, Reactor Engineering may , through a shutdown margin calculation, determine that the reactor will remain shutdown under all conditions without boron. If Reactor Engineering makes that determination , the condition to exit E0-000-113 is satisfied; the step may be answered " NO" and step LQ-3 may be performed. (Reference

PSTG override before C5-1; first override before RC/Q-1) LQ-3 STOP BORON INJECTION AND GO TO RPV CONTROL If it can be determined the reactor will remain shutdown under all conditions without boron , it is appropriate to return to the RPV Control procedure in order to restore and maintain RPV water level in its normal range. Boron injection is terminated and the procedure is exited. (

Reference:

PSTG override before C5-1; first overr i de before RC/Q-1)

LQ-4 ENSURE RX MODE SWITCH IN S/D E0-000-113 Revision 13 Page 6 of 67 When the reactor mode switch is placed in the SHUTDOWN position, a diverse and redundant reactor scram signal is generated by the RPS logic. Sets of contacts momentarily open which trip the RPS logic in the same manner as the automatic RPS logic trips. (

Reference:

PSTG RC/Q-1) LQ-5 ENSURE ARIINITIATED Initiation of ARI provides an independent and redundant means of depressurizing the reactor scram air header and operating the scram discharge volume vent and drain valves. (

Reference:

PSTG RC/Q-2) LQ-6 RECORD INITIAL ATWS PWR: _____________________

% The purpose of this step is to declare and record the " initial ATWS power level" for use in future decisions regarding RPV depressurization.

The 5% determination is based on initial ATWS power , or reactor power before any other steps to reduce reactor power are performed, that is, before boron is injected, reactor recirculation is tripped or RPV water level is lowered. There is no guarantee that RPV water level will remain low or boron will remain in the core following rapid depressurization

therefore, "initial ATWS power level" is recorded.

This value will be referenced by steps that direct actions to control power and perform rapid depressurization.

Loss of APRM indication, by itself, does not mean that reactor power cannot be determined.

The values of reactor period , steam flow, RPV pressure and pressure trend , number of open SRVs or main turbine BPVs , etc. are valid indications of reactor power level relative to this decision.

If reactor power cannot be determined by these methods for any reason , reactor power must be assumed to be above the APRM downscale trip and tripping reactor recirculation pumps is appropriate.

Fuel damaging, large amplitude power excursions may occur at low RPV pressures if reactor power is > 5%. Also, development of the HCTL assumes that the reactor is shutdown.

Therefore, depressurization due to HCTL or PSL during an ATWS is restricted to conditions where initial ATWS power level is ::; 5%. If power level cannot be determined or is oscillating around 5%, it is conservative to declare power> 5%.

E0-000-113 Revision 13 Page 7 of 67 Initial ATWS power does not change on re-entry to this procedure. (

Reference:

PSTG first override before RC/P-2, SP/T-3, PC/P-3) LQ-7 MONITOR AND CONTROL LQ/Q LQ/L AND LQ/P CONCURRENTLY The symptomatic approach to emergency response, upon which the EOPs are based , precludes being able to establish in advance a priority for executing any of the parallel action paths of Level/Power control. Rather, current values and trends of parameters and the status of plant systems and equipment dictate the relative importance of individual Level/Power Control steps and the relative priority with which they should be accomplished. (

Reference:

PSTG RC-1 Monitor and Control Concurrently)

LQ/Q -Reactor Power Control LQ/Q-1 LQ/Q-2 IS INITIAL ATWS PWR > 5% OR CANNOT BE DETERMINED An affirmative answer to this question results in immediate boron injection. When scram and ARI have failed , initial ATWS power must be assessed to determine if immediate boron injection is required. If initial ATWS power was greater than 5%, then a relatively large number of control rods have failed to insert. The seriousness of this condition requires immediate injection of boron to positively terminate the ATWS event. If power is oscillating around 5%, it is conservative to declare power is > 5%. Early boron injection has the following benefits:

• Stop or prevent large-magnitude Limit Cycle Oscillations which can lead to core damage.

• Limit fuel damage from uneven flux patterns that could result from partial rod inserts.

• Protect the primary containment from excessive heat input. (

Reference:

PSTG step RC/Q-2) BEFORE SUPP POOL TEMP REACHES 150°F GO TO LQ/Q-3 The intent of this step is to ensure , for a low power ATWS , that boron is injected early enough to minimize the challenge to primary containment integrity.

LQ/Q-3 E0-000-113 Revision 13 Page 8 of 67 If boron injection was not required due to initial power level and conditions have since changed, this step ensures boron initiation takes place early enough to minimize the challenge to primary containment integrity.

If boron injection is initiated before suppression pool temperature reaches 150°F, emergency RPV depressurization due to HCTL may be precluded for initial ATWS power levels s 5% for an isolated RPV with suppression pool cooling in service. At higher power levels, the HCTL will be exceeded.

This step will only be required if initial ATWS power was s 5%. If initial ATWS power was > 5%, boron injection would have been required per steps LQ/Q-3 or LQ/Q-4. The logic term " BEFORE" permits injection of boron at any temperature up to the suppression pool temperature of 150°F. If suppression pool temperature exceeds 150°F, the action of this step is still required.

Level/Power Control is entered from RPV Control if more than one control rod is > 00. If, however, a scram condition exists, reactor power iss 5%, and no other RPV control entry condition exists, there is no permission to enter Level/Power Control from the EOPs. To authorize use of alternate methods of control rod insertion in this situation, entry to Level/Power Control at this step is directed from ON-100-101 (ON-200-101)

Scram, Scram Imminent.

However, if any entry condition occurs while inserting control rods, E0-000-113 must be entered at LQ-1. (

Reference:

PSTG step RC/Q-6) INJECT SBLC AND INHIBIT ADS, IF NOT INITIATED ADS initiation may result in the injection of large amounts of relatively cold, unborated water from low pressure injection systems. With the reactor either critical or shutdown on boron, the positive reactivity addition due to boron dilution and temperature reduction through the injection of cold water may result in a reactor power excursion large enough to cause substantial core damage. Preventing ADS is therefore appropriate whenever boron injection is required. However, if ADS has already initiated prior to performing this step, it is not inhibited.

Actions to limit injection flow rates to maintain RPV level or flooded conditions will already be in progress, and the loss of boron will be minimized.

Additionally, inhibiting ADS if already initiated will conflict with direction provided by other Emergency Procedures.

Instructions for inhibiting ADS are located in OP-183-001 (OP-283-001

), Automatic Depressurization System & Safety/Relief Valves. (

Reference:

PSTG step RC/Q-2 and second override before RC/Q-1)

LQ/Q-4 LQ/Q-5 LQ/Q-6 IF BORON CANNOT BE INJECTED WITH SBLC INJECT BORON WITH RCIC lAW ES-150-002 (ES-250-002)

E0-000-113 Revision 13 Page 9 of 67 This step is applicable to all subsequent steps within this flowpath. It remains applicable to those steps until the flowchart is exited. If, at any time, it is determined that boron is not entering RPV, this step permits the use of the alternate injection method. ES-150-002 (ES-250-002), Boron Injection Using RCIC System , provides instructions for this alternate injection method. (

Reference:

PSTG second override before RC/Q-1) ENSURE: RWCU ISOLATED OR FIL TERIDEMINS BYPASSED AND ISOLATED Isolation of RWCU is desirable when boron is injected into the RPV because the demineralizers remove boron from the reactor coolant. However, RWCU may be restored and used to augment RPV pressure control lAW ES-161-002 (ES-261-002), RWCU Recirculation Mode Bypassing Interlocks.

For this condition, and when RWCU has failed to automatically isolate, operator action is to isolate the system. (Reference

PSTG second override before RC/Q-1) ENSURE SRM'S AND IRM'S INSERTED If all but a few control rods are inserted, reactor power will soon be downscale on APRMs and other means must be used to observe decreasing reactor power. Source range and intermediate range detectors are inserted to fulfill need for low power indication.

Inserting SRMs and IRMs is required regardless of reactor power, since they will function long enough to achieve reactor shutdown. (Reference

PSTG step RC/Q-2)

LQ/Q-7 LQ/Q-8 IF ANY STEAM TURBINE ON-LINE ENSURE RECIRC RUN BACK TO MIN E0-000-113 Revision 13 Page 10 of 67 An immediate and rapid reactor power reduction may be achieved by reducing reactor coolant recirculation flow. The most rapid flow rate reduction and, consequently, the most rapid power reduction, is achieved by tripping the recirculation pumps. However, if the recirculation pump trip is initiated from a high power level, the resulting rapid changes in steam flow, RPV pressure, and RPV water level may cause a trip of one or more of the following steam-driven turbines:

main turbine-generator, HPCI, RCIC and feedwater.

If the main turbine-generator trips and reactor power exceeds the turbine bypass valve capacity, RPV pressure will increase until one or more SRVs open. Heatup of the suppression pool then begins and boron injection may ultimately be required. If HPCI, RCIC , or feedwater trip, the resultant RPV water level transient may require emergency depressurization of the RPV and operation of less desirable RPV injection sources. To effect a more controlled reduction in reactor power and thereby avoid main turbine-generator and RPV injection system trips and their associated complications, a recirculation flow runback to minimum (::; 30% speed) is performed prior to tripping the recirculation pumps. The word " ensure" recognizes that an automatic runback may have already occurred. If it has, no other action is required.

If an automatic runback to the #1 limiter has not occurred, runback to minimum is accomplished by lowering the demand on the individual controllers to minimum. (

Reference:

PSTG step RC/Q-3) IF INITIAL ATWS PWR > 5% OR CANNOT BE DETERMINED TRIP BOTH RECIRC PUMPS A recirculation pump trip from high power provides a rapid reduction in power. If initial ATWS power was below the APRM downscale trip setpoint (5%), tripping the recirculation pumps results in little reduction in reactor power since power is already near the decay heat level. Diverse indications should be used to evaluate reactor power level. 3-D model tests have demonstrated that forced recirculation need not be maintained if boron injection is later required.

Natural circulation flow provides adequate boron mixing. (

Reference:

PSTG step RC/Q-4)

LQ/Q-9 LQ/Q-10 MAXIMIZE CRD E0-000-113 Revision 13 Page 11 of 67 This step is provided to increase cooling water differential pressure to maximum which may cause the control rods to drift inward and shutdown the reactor. Maximizing the unit's CRD system is a relatively fast method for inserting negative reactivity.

If the CRD system is not in service additional operator actions to restore the system is provided in step CR-4. (

Reference:

PSTG step RC/Q-7.2)

GO TO CONTROL ROD INSERTION This step divides reactor power control into two flowpaths.

Control rod insertion methods are given on sheet 2 of E0-000-113, and boron injection instructions are given in the Level/Power Control steps. Both flowpaths are performed concurrently as indicated by the exit-and-perform-concurrently arrow associated with this step. The symptom based approach to emergency response prevents assignment of priorities to either of these flowpaths since the time at which boron must be injected into the RPV is dependent on the magnitude of the ATWS event. This is a critical step if SLCS fails and the MSIVs are closed because primary containment venting can still be avoided if manual control rod insertion is initiated promptly.

For an MSIV-closure ATWS with complete failure to scram and complete failure of SLCS, reactor power can be reduced to within the capacity of the RHR system by individually driving control rods with the CRD system. (

Reference:

PSTG Execute Steps RC/Q-6 and RC/Q-7 Concurrently)

LQ/Q-11 IF SBLC TANK L VL DROPS TO 0 GAL STOP BORON INJECTION E0-000-113 Revision 13 Page 12 of 67 This step is applicable to all subsequent steps within this flowpath. It remains applicable to those steps until the flowchart is exited. Intent of this step is to preserve availability of the SBLC System should it be required for subsequent injection.

A tank level of 0 gallons ensures SBLC suction piping remains covered, so pump functionally is preserved.

OP-153-001, ( OP-253-001) and (ES-150-002 , ES-250-002) direct opening of heater supply breaker. The words "stop boron injection" are intentionally used rather than " trip SBLC pumps" to recognize that either SBLC pumps or RCIC, if aligned for alternate boron injection , may be accomplishing boron injection.

An indicated tank level of 0 gallons means that a volume of boron greater than the Cold Shutdown Boron Weight (CSBW) 1350 gallons of 10% concentration, has been injected.

CSBW is defined to the least amount of soluble boron which , if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions. If the SBLC system functions are designed and SBLC tank level starts at greater than or equal to the minimum required by Technical Specifications, a volume greater than CSBW will be injected when tank level decreases to 0 gallons. CSBW is determined assuming:

1. All rods are full out. 2. Core is at most reactive exposure. 3. No Xe in core. 4. No voids in core. 5. Water temperature 68°F. 6. Shutdown cooling and RWCU is in service. 7. Reactor level is at +54". (

Reference:

PSTG second override before RC/Q-1)

E0-000-113 Revision 13 Page 13 of 67 LQ/L -RPV Level Control LQ/L-1 ENSURE ALL:

• ISOLATIONS

• ECCS INITIATIONS

• DG'S START Intent of this step is to quickly assess plant status and to determine proper automatic operation of plant equipment occurred.

SPDS may be used to determine Containment Isolations. "Ensure" means take manual action for any automatic operation that should have occurred but did not. Diesel generator initiation assures that there is redundant source of electrical power available for RPV water level control. A loaded diesel generator must be supplied with adequate ESW flow within 4'Y:! minutes. This limit is extended to 8 minutes if diesel generator is running unloaded.

Adequate ESW flow is describep in OP-054-001, Emergency Service Water System. Instructions on how to manually shut-down a running diesel generator are located in OP-024-001.

Instructions to bypass interlocks lAW ES procedures always supersede this step's requirements. (

Reference:

PSTG RC/L-1)

LQ/L-2 IF WATER LVL CANNOT BE DETERMINED GO TO RPV FLOODING CAUTION E0-000-113 Revision 13 Page 14 of 67 1 RPV WATER LVL INSTR MAY BE USED WHEN DETERMINED USABLE lAW ON-145-004 (ON-245-004)

OR IT READS WITHIN BAND (INCHES):

SHUTDOWN 50-500 UPSET 45 .. 180 EXTENDED 180 NARROW 2 .60 WIDE NO FIRE ON 749 -145 60 X218_Z7 (X232_Z5)

ALARM -1 60 FUEL ZONE -290 -110 This step is applicable to all subsequent steps within this flowpath.

It is applicable whether the step was entered from above , or transfer resulted in entry below this step. It remains applicable to those steps until flowchart is ex i ted. As documented in RC/L-2 , off-scale RPV water level indication alone is not an indication that " RPV water level cannot be determined

". The decision is made based on knowing whether or not level can be determined. If level indication is believed to be erroneous or unbel i evable, entry to E0-000-114, RPV FLOODING, is warranted.

If at any time during the performance of this procedure, RPV water level cannot be determined,.RPV Flooding is required.

Caution 1 is applicable throughout this flowpath.

LQ/L-3 LQ/L-4 LQ/L-5 E0-000-113 Revision 13 Page 15 of 67 Factors inherent in the design of the RPV water level measurement systems make the validity of information supplied by these instruments dependent on the value of RPV pressure, drywell temperature, and reactor building temperature.

The above caution provides the minimum usable water levels when drywell and reactor building temperatures near these instrument reference leg vertical runs are at their bounding values. ON-145-004 (ON-245-004), RPV Water Level Anomaly, identifies the specific conditions in which the instruments are usable. (See E0-000-100 for discussion of CAUTION 1.) (

Reference:

PSTG override before C5-1) RESET MAIN GEN LOCKOUTS The intent of this step is to prevent an inappropriate auxiliary bus load shed. Since spurious operation of the load shed scheme will cause an undesirable loss of loads such as condensate pumps, service water pumps, circulating water

• pumps, auxiliary boilers and turbine building heating load centers , main generator lockouts are reset if RPV level can be maintained above -129". (Reference

PSTG C5-2) IS INITIAL ATWS PWR > 5% OR CANNOT BE DETERMINED Answer to this question will determine whether RPV water level should be lowered, or whether it should be maintained in the normal operating band. (

Reference:

PSTG step C5-3) INHIBIT ADS , IF NOT INITIATED Actions in this flowpath lower RPV water level near the automatic initiation setpoint of ADS. Rapid and uncontrolled injection of relatively cold, unborated water from low pressure injection systems may occur as RPV pressure decreases below the shutoff heads of these pumps. This would quickly dilute boron concentration in the core and reduce reactor coolant temperature.

When the reactor is not shutdown, or when the shutdown margin is small , sufficient positive reactivity might be added in this way to cause a reactor power excursion.

Therefore , ADS initiation is purposely prevented as one of the first actions. When required to initiate ADS, explicit direction to depressurize the RPV is provided in the EOPs. However , if ADS has already initiated prior to performing this step, it is not inhibited.

Actions to limit injection flow rates to maintain RPV level or flooded conditions will already be in progress, and the loss of boron will be minimized. Additionally, inhibiting ADS if already initiated will conflict with direction provided by other Emergency Procedures. PSTG C5-3)

LQ/L-6 USING TABLE 15 SYSTEMS IRRESPECTIVE OF LVL AND PWR OSCILLATIONS THROTTLE INJECTION AND PREVENT INJECTION UNTIL LVL BETWEEN -60" AND -110" E0-00.0-113 Revis i on 13 Page 16 of 67 The purpose of this step is to uncover the feedwater spargers sufficiently to reduce core inlet subcooling. A General Electric Company study (NED0-32047) indicates that the major threat to fuel integrity from ATWS is caused by large-amplitude power/flow instabilities. These density-wave instabilities will most likely develop in the non-isolation ATWS where the feedwater system is still available for makeup to the RPV. In this event , the feedwater system maintains normal water level, but feedwater

• heating is lost due to tripping of the turbine. Without preheating of the feedwater , high levels of core-inlet subcooling develop which can drive the reactor into a highly unstable mode of operation.

General Electric calculations indicate that power oscillations become large enough to cause melting of fuel in high-power bundles. In the non-isolation ATWS event, clad damage due to unstable operation can be prevented or at least mitigated by promptly reducing feedwater flow so that level is lowered below the feedwater spargers.

Once level drops below the sparger nozzles, which are located at -24" , the feedwater is sprayed into a region occupied by saturated steam. Steam will then condense on the injected feedwater , and the coolant will be heated as it falls to the liquid surface within the downcomer.

Heating of the feedwater by steam condensation limits the buildup of core inlet subcooling and can prevent the onset of severe power/flow instabilities.

E0-000-113 Revision 13 Page 17 of 67 Although the purpose of this step is to uncover the feedwater spargers sufficiently to reduce inlet subcooling, water level will eventually be maintained between -161" and -60" per step LQ/L-13 to effect a reduction in reactor power, thereby reducing heat rate to the containment.

• SLC

• FW

• COND TABLE 15 SYSTEMS

• CRD MAXIMIZED AS NECESSARY

• RCIC WITH SUCTION FROM CST IF AVAILABLE

• HPCI WITH SUCTION FROM CST IF AVAILABLE

• LPCI WITH INJECTION THRU HX ASAP CAUTIONS 140 OPERATION OF HPCI OR RCIC WITH SUCTION FROM SUPP POOL AND SUPP POOL TEMP > 140°F MAY RESULT IN EQUIPMENT DAMAGE. 50 ELEVATED SUPP CHMBR PRESS MAY TRIP RCIC ON HIGH EXHAUST PRESS. 11 HPCI INJECTIONS CAUSE THERMAL SHOCK TO RPV AT FEEDWATER PENETRATIONS. VL OPERATION OF RHR OR CORE SPRAY WITH SUCTION FROM SUPP POOL AND PUMP FLOW BELOW VL MAY RESULT IN EQUIPMENT DAMAGE. The preferred systems for use in controlling RPV water level are those Table 15 Systems which inject into the feedwater sparger or outside the core shroud. These are used because cold water is preheated by steam and the flow path outside the core shroud mixes the relatively cold injected water with the warmer water in the lower plenum prior to reaching the core. Injection from SLC and CRD are always permitted during ATWS events. The operator throttles existing injection except CRD and SLC and prevents unwanted injection as necessary to decrease level. Preventing injection such as RCIC and HPCI as level drops below -30" and -38" respectively may be required when feedwater is available.

LQ/L-7 LQ/L-8 E0-000-113 Revision 13 Page 18 of 67 LPCI is aligned with injection through the heat exchanger as soon as possible. The benefit of injecting through the heat exchanger is that it results in containment heat removal. There is very little increase in core inlet subcooling with the heat exchangers aligned , so the risk of unstable operation is not increased appreciably. (Reference

PSTG C5-3) BYPASS RCIC LOW PRESS ISO AS NECESSARY lAW ES-150-001 (ES-250-001)

OPERATE RCIC AS NECESSARY lAW ES-150-003 (ES-250-003)

SUPPLYING 125VDC LOADS AS NECESSARY lAW ES-002-001 When necessary , low RPV pressure isolation interlocks are defeated for RCIC because some injection into the RPV can still be sustained above the turbine stall pressure. However, the higher steam demand required for HPCI operation prevents sustained operation at low pressure.

During a loss of AC power to 125VDC battery chargers 1 D613, 2D613, 1 D623 and/or 2D623, long term availability of HPCI, RCIC and/or ADS beyond 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requires use of the portable generator (OG503) to maintain 125VDC control power. (

Reference:

PSTG C5-3) IF ANY MSL IS NOT ISOLATED BYPASS MSIV AND CIG INTERLOCKS RPV water level is lowered to near the MSIV isolation set point , so it is appropriate to prevent MSIV isolation to preserve the Main condenser as a heat sink as long as possible. CIG isolation interlocks are bypassed to ensure an uninterrupted pneumatic supply pressure to hold open the MSIVs. OP-184-001 (OP-284-001

), Main Steam System, provides instructions for bypassing the MSIV and CIG interlocks. (

Reference:

PSTG C5-3)

LQ/L-9 CAUTION LOSS OF 1-A E0-000-113 Revision 13 Page 19 of 67 CAUSES LOSS OF NORMAL FW INJECTION Purpose of this caution is to advise the operator of the undesirable results from failure to promptly recover from the loss of instrument air. It is recognized that instrument air can be lost for many reasons, but in Level/Power Control, the probability of losing instrument air is increased since level is intentionally lowered close to -129" and containment pressure is likely increasing close to 1.72 psig. (

Reference:

PSTG Caution 9) IF PC PRESS CANNOT BE MAINTAINED<

65 PSIG STOP RPV INJECTION FROMSOURCESEXTERNALTOPC EXCEPT FROM:

• SYSTEMS NEEDED FOR ADEQUATE CORE COOLING

• SBLC

• CRD

• RCIC This step is applicable to all subsequent steps within this flowpath.

It remains applicable to those steps until flowchart is exited. If an unisolable break exists inside the drywell, continued RPV injection from sources external to the primary containment will increase primary containment water level after RPV water level reaches the elevation of the break. The increasing primary containment water level will, in turn, increase the hydrostatic pressure over submerged components and compress the primary containment airspace, thereby increasing the atmospheric pressure.

Above 65 psig, primary containment vent valve operability is not assured. Vent valves are located in both the drywell and suppression chamber, therefore, primary containment pressure must be monitored in both compartments.

If primary containment pressure, either drywell or suppression chamber, cannot be maintained below 65 psig, injection into the RPV from sources external to primary containment is terminated unless injection is needed for adequate core cooling. If adequate core cooling cannot be assured, injection from sources external to the primary containment is continued.

LQ/L-10 LQ/L-11 E0-000-113 Revision 13 Page 20 of 67 This override terminates only injection

" from sources external to the primary containment." Injection from the suppression pool may continue.

Injection from boron injection systems and CRD is not terminated because operation of these systems may be needed to establish and maintain the reactor shutdown. (

Reference:

PSTG override before C5-1) IF RAPID DEPRESS REQ'D GO TO LQ/L-18 This step is applicable to all subsequent steps within this flowpath. It remains applicable to those steps until flowchart is exited. The steps which follow specify the use of various systems to control RPV water level. If rapid depressurization of the RPV is required, these systems must be operated to minimize the potential for rapid injection of cold, unborated water into the core region as RPV pressure decreases below pump shutoff head. Steps following LQ/L-18 provide appropriate instructions for controlling injection systems in this manner. (Reference

PSTG 1st override before C5-5) IS INITIAL ATWS PWR > 5% OR CANNOT BE DETERMINED Answer to this question determines which water level control band will be prescribed. (Reference

PSTG step C5-5) CAUTION A RAPID INCREASE IN INJECTION MAY INDUCE LARGE PWR EXCURSION AND SUBSTANTIAL CORE DAMAGE. This caution is applicable throughout this flowpath. This Caution highlights the potential for large reactor power excursions and subsequent core damage from rapid injection of cold, unborated water when injection is continued , per the next step, to maintain RPV water level within the prescribed band. (

Reference:

PSTG Caution 6)

LQ/L-12 MAINTAIN LVL BETWEEN -129" AND +54" USING TABLE 15 SYSTEMS E0-000-113 Revision 13 Page 21 of 67 BYPASSING RCIC LOW PRESS ISO AS NECESSARY lAW ES-150-001 (ES-250-001)

OPERATE RCIC AS NECESSARY lAW ES-150-003 (ES-250-003).

SUPPLYING 125VDC LOADS AS NECESSARY lAW ES-002-001.

A wide water level control band is given to afford maximum flexibility in controlling water level. If water level had been controlled in the normal operating range and initial ATWS power level 5%, the wide band allows level to be maintained in the normal band of +13" to +54". If the initiating event was a low RPV water level condition, it cannot be known if the low water level had been the reason that initial power level 5%. The wide water level control band allows level to be controlled at the level it bottomed out at, but above -129". This lower limit is intended to remove any incentive to recover level to the normal operating band if it is possible that low water level had suppressed the initial ATWS power level. Table 15 systems prescribed here are the same systems prescribed in LQ/L-6. Refer to that step for a complete explanation of the systems and cautions applicable to their use. The basis for defeating the RCIC system interlocks and supplying 125VDC loads is given in step LQ/L-7. (

Reference:

PSTG C5-5) CAUTION A RAPID INCREASE IN INJECTION MAY INDUCE LARGE PWR EXCURSION AND SUBSTANTIAL CORE DAMAGE This caution is applicable throughout this flowpath. This Caution highlights the potential for large reactor power excursions and subsequent core damage from rap i d injection of cold , unborated water when injection is continued, per the next step, to maintain RPV water level within the prescribed band. (

Reference:

PSTG Caution 6)

LQ/L-13 MAINTAIN LVL BETWEEN -60" AND -161" USING TABLE 15 SYSTEMS E0-000-113 Revision 13 Page 22 of 67 BYPASSING RCIC LOW PRESS ISO AS NECESSARY lAW ES-150-001 (ES-250-001)

OPERATE RCIC AS NECESSARY lAW ES-150-003 (ES-250-003).

SUPPLYING 125VDC LOADS AS NECESSARY lAW ES-002-001.

This step identifies the widest, acceptable water level control band. Although level fluctuations within this band are safe , it is very desirable to maintain level within the more restrictive target area of -11 0" to -60". The target area and expanded band are shown in Figure 8, Water Level Operation Guidance.

The intent of this step is to remain within the target band at all times unless prohibited by system perturbations, and remain within the expanded band at all times. Operation outside the target area has the following disadvantages

The basis for an upper level of -60" is given in LQ/L-6. A lower level of -11 0" is specified for the following reasons: 1. Provides a margin for core coverage. 2. Avoids operation near TAF where core power is more responsive to RPV pressure fluctuations.

3. Makes level control easier by maintaining level above the narrow region of the downcomer.

Below -11 0" the downcomer free area reduces from 300 fe to 88 fe resulting in increased magnitude of indicated level oscillations.

4. Maintains sufficient core flow to carry liquid boron from lower plenum upward into the core. As level is decreased below -11 0" , boron mixing efficiency is reduced because the natural circulation flow rate through the jet pumps is reduced and not as efficient at carrying the injected boron from the lower plenum upward into the core. At very low downcomer water levels near or below top of active fuel, there is little water available in the region above the jet pump throat for mixing with boron injected via RCIC. In this situation, there is concern that boron may accumulate in the stagnant region of the downcomer which is below the jet pump throat.

E0-000-113 Revision 13 Page 23 of 67 5. Water level can be determined from wide range level instrumentation. 6. Avoids MSIV isolation setpoint of -129". CAUTION 12 PROLONGED OPERATION IN YELLOW AREA OF FIG 8 MAY RESULT IN ADDITIONAL CONTAINMENT LOADING AND PWR INSTABILITIES.

LQ/L-14 This caution is added to alert the operator of the undesirable affects of operating outside the target band of -11 0" to -60". Operation* outside the target band can result in increased power level, increased containment loading , and the potential for power oscillations.

Table 15 systems prescribed here are the same systems prescribed in LQ/L-6. Refer to that step for a complete explanation of the systems and cautions applicable to their use. If this step is entered following Rapid Depressurization and systems listed in LQ/L-21 are being used, their continued use is allowed. The basis for defeating the RCIC system interlocks and supplying 125VDC loads is given in Step LQ/L-7. (Reference

PSTG C5-5) IF L VL CANNOT BE RESTORED AND MAINTAINED>

-161" 1 GO TO LQ/L-18 2 GO TO RAPID DEPRESS This step is applicable to all subsequent steps within this flowpath.

It remains applicable to those steps until flowchart is exited. The intent of this step is to specify the limit when rapid depressurization of the RPV is appropriate despite the possibility of creating power/flow instabilities at low pressure.

E0-000-113 Revision 13 Page 24 of 67 Depressurizing a critical core results in destabilizing effects and has a potential to cause core damage. The initial influence of reactor depressurization is stabilizing since the additional flashing of liquid phase required for depressurization introduces excess voids in the reactor core which can essentially terminate the fission process if the rate of depressurization is high enough. Once the depressurization is complete, however, the result is the immediate initiation of power excursions. Core damage is expected to occur from high clad stresses induced by: temperature excursions above the rewet temperature, PCI, cyclic fatigue, burnout or having the fuel enthalpy exceed the cladding failure threshold.

Core destabilizing effects are mitigated by boron injection.

Therefore, the decision to perform rapid depressurization must not be made too early since an earlier RD results in less boron being present in the RPV when the RD is taken. While the goal is to perform the RD as close to -161 11 as possible, the wording of the step gives flexibility to perform the action after reaching -161 II. The determination that level cannot be restored and maintained

> -161 II must be based upon:

• availability of high pressure injection systems, and

• present level trend For example, level may have dropped below -161 11 , but the level trend shows that it will be able to be recovered above the limit. In this case, rapid depressurization should be deferred.

Controlling reactor pressure, power and level with condensate and SRVs at 500 psig is difficult because all 3 parameters affect each other. Therefore, rapid depressurization is recommended when high-pressure injection cannot be obtained. (Reference

PSTG C5-5 and C5-6)

LQ/L-15 LQ/L-16 IF PWR COMMENCES AND CONTINUES TO RAISE GO TO LQ/L-5 E0-000-113 Revision 13 Page 25 of 67 This step is applicable to all subsequent steps within this flowpath. It remains applicable to those steps until flowchart is exited. If reactor power commences and continues to increase as RPV water level is raised, the amount of boron required to shut down the reactor (HSBV'{) has not reached the core. The return to step LQ/L-5 will again require that RPV water level be lowered to suppress reactor power while additional actions to shut down the reactor proceed. As injection into the RPV is initially increased to raise RPV water level, a small transient increase in reactor power is expected as natural circulation core flow is re-established.

This power increase will be reversed as boron i s mixed and carried from the lower plenum up into the core region. The wording in the above step , " ... commences and continues to increase ... ," denotes only a sustained increase in reactor power indicative of insufficient boron in the core. (

Reference:

PSTG override preceding C5-6) WHEN HOT S/D BORON WEIGHT IS INJECTED TANK LVL Table 19. RESTORE AND MAINTAIN LVL BETWEEN +13" AND +54" Hot Shutdown Boron Weight (HSBW) is defined to be the weight of soluble boron which, if injected into the RPV and uniformly mixed, will maintain the reactor shutdown under hot standby conditions.

It assures the reactor will be shutdown irrespective of control rod position when RPV water level is raised to uniformly mix the injected boron. When HSBW has been injected , RPV water level is restored to and maintained within the normal operating range. If SLC system functions as designed and SLC tank level starts at or greater than the minimum required by Technical Specifications, then HSBW will be injected when tank level decreases to Table 19 gallons. The tank concentration is assumed to be at the Technical Specification minimum concentration.

This takes about 25 minutes. As RPV water level is increased, natural circulation flow is increased and the boron which may have accumulated in the lower plenum is quickly mixed and distributed throughout the core region. Delays in raising RPV water level increases the time the core may be at power , and may result in a greater challenge to containment.

The HSBW is determined assuming:

E0-000-113 Revision 13 Page 26 of 67 1. Control rods withdrawn to maximum rod block limit. 2. Core is at most reactive exposure. 3. Hot full power Xe. 4. No voids in core. 5. RPV water temperature is the saturation temperature for the lowest lifting SRV setpoint pressure.

6. Water level is at the high level trip setpoint (+54 in.). 7. No shutdown cooling is in service. 8. RWCU is in service. The intent of this step is to wait until HSBW is injected before raising RPV level. In those circumstances when boron does not reach the RPV and is lost for this purpose , this step requires that HSBW be injected even if this requires refilling SLC tank with new boron. RPV water level is restored and maintained with the systems which were effectively providing RPV injection in the previous steps of this flowchart.

For example: if RPV water level could not be restored until Core Spray was placed in service in step LQ/L-21 , the continued use of Core Spray is appropriate in this step. TABLE 19 HSBW INJECTED INITIAL TANK FINAL TANK VOLUME VOLUME 2000 1150 1900 1060 1800 975 1700 891 1600 806 1500 722 1400 631 (Reference

PSTG C5-6)

LQ/L-17 LQ/L-18 WHEN L VL CANNOT BE RESTORED AND MAINTAINED>

+13" MAINTAIN LVL > -161" E0-000-113 Revision 13 Page 27 of 67 If RPV water level cannot be maintained in the normal range, an alternate control band with a lower limit is defined. The widened RPV water level control band provides added operational flexibility while still assuring adequate core cooling through core submergence.

It makes available additional time to place in service those injection systems not yet operating.

The widened band also accommodates controlling RPV water level , without employing additional contingency actions, for a condition where a break exists between the low RPV water level scram setpoint and TAF where injection flow cannot overcome break flow. (Reference

PSTG C5-6) STOP INJECTION AND PREVENT INJECTION FROM:

• FW

• COND

• LPCI

• CORE SPRAY

• HPCI Intent of this step is to prevent uncontrolled injection of large amounts of cold water as RPV pressure decreases below the shutoff head of operating system pumps. For feedwater, this would mean tripping feedwater pumps or closing their discharge valves. For condensate, this would mean preventing injection below RPV pressure of 700 psig using values or if needed, tripping condensate pumps. For LPCI and Core Spray, this would require preventing injection in accordance with overriding section of their respective operating procedures.

For HPCI, this would require preventing injection in accordance with TERMINATE HPCI DURING AN ATWS section of HPCI operating procedures.

E0-000-113 Revision 13 Page 28 of 67 Injection from boron injection systems and CRD is not terminated because operation of these systems may be needed to establish and maintain reactor shutdown.

Further, the injection flowrates from these systems are small compared to those of the other Table 15 systems. Injection from RCIC is not stopped because the injection flowrate from this system is small. (

Reference:

PSTG C5-5.1) CAUTION A RAPID INCREASE IN INJECTION MAY INDUCE LARGE PWR EXCURSION AND SUBSTANTIAL CORE DAMAGE. This caution is applicable throughout this flowpath.

The applicability of this Caution highlights the potential for large reactor power excursions and subsequent core damage from rapid injection of cold, unborated water when injection is commenced after depressurization is initiated. (

Reference:

PSTG Caution 6)

LQ/L-19 WHEN RAPID DEPRESS HAS BEEN INITIATED COMMENCE AND IRRESPECTIVE OF VORTEX LIMITS SLOWLY RAISE INJECTION TO RESTORE AND MAINTAIN L VL BETWEEN -60" AND -161" USING TABLE 15 SYSTEMS E0-000-113 Revision 13 Page 29 of 67 Intent of this step is to re-establish injection in a controlled manner after rapid depressurization has been initiated. "Initiated" is defined as ADS valves have been opened, either automatically or manually.

It is not intended that the depressurization process be completed , only initiated.

As long as any number of ADS valves have been opened in response to an automatic signal or manual action , this condition is met and injection may be slowly re-established. Steps LQ/L-6 , LQ/L-12 and LQ/L-13 permit use of these same Table 15 systems. Refer to RC/L-6 for a complete explanation of the systems and cautions applicable to their use. Here , however , an explicit direction is given to commence injection

" irrespective of vortex limits ," since restoration of adequate core cooling takes precedence over adherence to normal operating limits. The undesirable consequences of uncovering the reactor core outweigh the risk of equipment damage which could result if vortex limits are exceeded. Immediate and catastrophic pump failure is not expected to occur should operation beyond these limits be required. A specific order governing the priority over use of these systems cannot be predetermined as it will depend greatly on plant conditions. Consider the following factors to determine order:

• Injection through FW spargers (preferred)

• . System availability

• Time and manpower required to operate system

• System throttling capability/control

• Water quality Re-establishing injection into the RPV is required in order to adequately cool the core and to make up the mass of steam being rejected through open SRVs. Since the reactor may become critical during this evolution, injection into the RPV is increased slowly to preclude the possibility of large power excursions caused by rapid injection of cold unborated water.

E0-000-113 Revision 13 Page 30 of 67 The level control band of -60" to -161" is the widest, acceptable water level control band. Although level fluctuations within this band are safe, it is very desirable to maintain level within the more restrictive target area of -11 0" to -60". The target area and expanded band are shown in Figure 8, Water Level Operation Guidance.

Operation outside of the target area has numerous disadvantages which are described in LQ/L-13. The intent of this step is to restore and maintain level within the target band at all times unless prohibited by system perturbations or inadequate vessel injection, and remain within the expanded band at all times. CAUTION 12 PROLONGED OPERATION IN YELLOW AREA OF FIG 8 MAY RESULT IN ADDITIONAL CONTAINMENT LOADING AND PWR INSTABILITIES.

LQ/L-20 This caution is added to alert the operator of the undesirable affects of operating outside the target band of -11 0" to -60". Operation outside the target band can result in increased power level, increased containment loading, and the potential for power oscillations. (

Reference:

PSTG C5-5.2) CAN LVL BE RESTORED AND MAINTAINED>

-161" This decision tests the success of the previous actions to maintain the core submerged. (

Reference:

PSTG C5-5.2) CAUTION A RAPID INCREASE IN INJECTION MAY INDUCE LARGE PWR EXCURSION AND SUBSTANTIAL CORE DAMAGE. This caution is applicable throughout this flowpath. The applicability of this Caution highlights the potential for large reactor power excursions and subsequent core damage from rapid injection of cold, unborated water when injection is commenced. (

Reference:

PSTG Caution 6)

LQ/L-21 COMMENCE AND IRRESPECTIVE OF VORTEX LIMITS SLOWLY RAISE INJECTION TO RESTORFAND MAINTAIN LVL > -161" USING ANY:

• CORE SPRAY

• RHRSW X-TIE FROM EITHER UNIT

• FIRE SYSTEM lAW ES-013-001

• FIRE SYSTEM THROUGH RHR A KEEPFILL

• CRD X-TIE TO OTHER UNIT

• ECCS KEEP-FILL

• RHR SOC SUCTION FILL E0-000-113 Revision 13 Page 31 of 67

• PORTABLE PUMP TRUCK lAW DC-B5B-102 (DC-B5B-202)

Same as LQ/L-19, vortex limits do not apply. If RPV level cannot be maintained above -161" using the preferred systems , other systems are required.

Core Spray injects inside the shroud. RHRSW injects low quality water. ECCS Keep-Fill and RHR SOC Suction Fill are a relatively low capacity source of injection.

SLC is not included in this list. The ATWS event requires that SLC should remain available for the injection of boron by the reactor power control flowpath.

A specific order governing the priority over use of these systems cannot be predetermined as it will depend greatly on plant conditions.

Consider the following factors to determine order:

• System availability

• Time and manpower required to operate system

• System throttling capability/control

• Water quality Alternate injection sources under failure-to-scram conditions include those that inject inside the core shroud and alternate injection subsystems which are those that inject water of a lower quality , those that require unusual or complex lineups , and those that would be u*sed only if a preferred system were unavailable. Per BWROG guidance , some injection methods may be appropriate for consideration under severe accident conditions when additional support personnel are available but impractical for use under emergency conditions by the normal shift complement.

Use of the Portable Pump Truck for injection per the Damage Control procedure is not considered practical for the on-shift Operations crew without additional support during EOP implementation.

LQ/L-22 E0-000-113 Revision 13 Page 32 of 67 The Fire System has the capability of being lined-up by two flow paths. The injection path using the RHR A Loop Keep-fill system is less complex if access to the Reactor Building exists, but has a lower injection capability (approx. 300 gpm at 0 psig RPV). The line-up through the RHRSW system is more complex, but has the greater injection capability (approx. 1100 gpm at 0 psig RPV). Both paths can be aligned simultaneously.

It is expected that injection through the RHR A Loop Keep-fill path would be established first. Fire pump injection to the vessel should then be maximized by follow-up actions to align the RHRSW path and inject through both paths simultaneously.

Instructions for aligning these systems are given in the following procedures:

• Core Spray OP-151-00 1 ( OP-251-00

1) Core Spray System

• RHRSW Cross-Tie From Either Unit OP-116-001 (OP-216-001)

RHR Service Water System

• Fire System ES-013-001 Fire Protection System Cross-Tie to RHRSW and OP-14.9-001 (OP-249-001)

RHR System

• CRD Cross-Tie to Other Unit OP-155-001 (OP-255-001)

Control Rod Drive Hydraulics System

• ECCS Keep-Fill OP-149-00 1 ( OP-249-00

1) RH R System

• RHR Suction Fill OP-149-002 (OP-249-002)

RHR Shutdown Cooling

• Portable Pump Truck DC-B5B-102 (DC-B5B-202)

Connection Of Portable Pump Truck To The RHRSW System To Provide Alternate Means Of Low Pressure RPV Injection, Containment Sprays Or Suppression Pool Makeup To Unit One (Two) (Reference

PSTG C5-5.2) CAN L VL BE RESTORED AND MAINTAINED>

-161" If level can be restored and maintained above -161", control is transferred back to LQ/L-13 where a new level band is given.

LQ/L-23 CONTINUE RPV INJECTION AND CONTACT TSC TO ENTER EP-DS-002 RPV AND PC FLOODING E0-000-113 Revision 13 Page 33 of 67 When injection with all available sources fails to provide sufficient injection to establish and maintain adequate core cooling, as a last resort , Level/Power Control water level actions are abandoned and submergence of the core is attempted through flooding the primary containment.

Control of RPV pressure remains in E0-000-112 and control of RPV water level is transferred to EP-DS-002 , RPV and Primary Containment Flooding. RPV and Primary Containment Flooding continues injec t ion with those systems authorized in Level/Power Control. The only difference of i njection sources is RPV and Primary Containment Flooding authorizes alignment of core spray suction to the CST , thereby increasing the contribution from outside containment sources. The additional contribution from the CSTs (and RWST if cross-tied) is insignificant compared to the volume required to fill primary containment to TAF. While flooding primary containment, RPV and primary containment venting are required when water level reaches 85ft. Therefore , there is no minimum time limit specified for making the decision to enter RPV and Primary Containment Flooding.

Alternate injection subsystems specified in LQ/L-21 may have become available if rapid depressurization took place at TAF. Depending on power level and which systems are presently injecting or expected to become available , the time to recover level to TAF will vary. Unlike RPV Control where the reactor is shutdown , a general guideline for determining when transfer to RPV and Primary Containment Flooding is appropriate cannot be given. (

Reference:

PSTG override preceding C5-5.3; step C5-5.3) LQ/P -RPV Pressure Control LQ/P-1 BEFORE DEPRESSURIZING<

700 PSIG PREVENT UNCONTROLLED COND INJECTION This step requires that action be taken prior to depressurizing the RPV to less than 700 psig which is the shutoff head of the condensate pumps. If the RPV is depressurized t o less than 700 psig w i thout the RFP discharge valves closed and the condensate system is in service, then an uncontrolled flood of the RPV will take place as condensate water injects through the feed pumps into the reactor. (

Reference:

PSTG RC/P)

LQ/P-2 IF E0-000-113 Revision 13 Page 34 of 67 ECCS INITIATION IMMINENT BEFORE DEPRESSURIZING

< 400 PSIG PREVENT INJECTION FROM LPCI AND CS PUMPS NOT REQ'D TO ASSURE ADEQUATE CORE COOLING This step is applicable to all subsequent steps within this flowpath.

It remains applicable to those steps until flowchart is exited. The intent of this step is to prevent injection that would complicate efforts to control RPV water level. The ECCS initiation signals for Susquehanna include either 1) low RPV level, or 2) a combination of high drywell pressure and low RPV pressure.

The word "imminent" is used to compensate for the combined initiation signal, since the existence of high drywell pressure and low RPV pressure would result in an ECCS injection, the very thing this step is trying to prevent. By requiring injection to be overridden when ECCS initiation is imminent, injection can be overridden when either signal exists (high drywell pressure or low RPV pressure), thereby, preventing ECCS injection before it occurs. The term "prevent" permits securing pumps to preclude injection.

The subsequent use of these systems is not prohibited by this override statement when plant conditions change such that system operation is required to assure adequate core cooling. Discussion of "prevent injection" is located in LQ/L-18. (

Reference:

PSTG override before RC/P-1)

LQ/P-3 IF ANY SRV IS CYCLING MANUALLY OPEN SRV'S UNTIL PRESS DROPS TO 945 PSIG E0-000-113 Revision 13 Page 35 of 67 "SRV cycling" is defined as multiple, closely sequenced valve actuations with valve opening being initiated in response to RPV pressure increasing to/above the lifting setpoint, valve closure being governed by RPV pressure decreasing to/below the reset setpoint.

The consequences associated with SRV cycling require prompt manual action to reduce RPV pressure below the SRV lifting setpoint.

Actions to prevent SRV cycling will minimize:

• Significant dynamic loads/stresses imposed on the RPV , on the SRV tail pipes and supporting structures , and on the primary containment structures.

• Fluctuating RPV water level (shrink occurring when the valves close as RPV pressure starts once again to increase , and swell occurring when the valves open as RPV pressure rapidly decreases).

• Repeated challenges to SRV operability (the potential failure of a valve to open on demand or to close once it has opened). SRV cycling is terminated by manual action to reduce RPV pressure to substantially below 1106 psig (the lowest SRV lifting setpoint). Manual operation of the SRVs to effect the desired prompt reduction in RPV pressure has the following advantages

• Level swells will be minimized because the SRVs cycle in groups in automatic but in manual control one SRV at a time may be used.

• Operable irrespective of MSIV status.

• Magnitude of RPV depressurization can be directly controlled.

RPV pressure reduction with SRVs is continued until RPV pressure reaches the pressure at which steam flow through the main turbine bypass valves is at 100% of bypass valve capacity.

If the MSIVs are open, reducing RPV pressure to below this value results in partial closure of the bypass valves and a corresponding increase in the amount of steam discharged to the suppression pool through the SRVs. If the MSIVs are not open, reducing RPV pressure to the lowest pressure at which all turbine bypass valves would be fully open if controlling pressure provides an adequate operating margin below the setpoint pressure of the lowest lifting SRV.

LQ/P-4 E0-000-113 Revision 13 Page 36 of 67 Since a prompt reduction in RPV pressure is desired as soon as possible to discontinue SRV cycling , adherence to a specific SRV opening sequence is unwarranted in this step. (Reference

PSTG step RC/P-1) IF INITIAL ATWS PWR 5% AND SUPP POOL TEMP AND LVL CANNOT BE MAINTAINED BELOW FIG 2 HCTL MAINTAIN RPV PRESS BELOW LIMIT EXCEEDING COOLDOWN RATE IF NECESSARY This step is applicable to all subsequent steps within this flowpath. It remains applicable to those steps until flowchart is exited. If suppression pool temperature and level cannot be maintained below HCTL (Heat Capacity Temperature Limit) the primary containment vent valve opening pressure may be exceeded following RPV depressurization. Refer to step SP/T-4 in E0-000-103 for discussion of HCTL. The figure is not provided on this flowchart because the condition requires that containment parameters be controlled first. The containment control procedure will provide the required figure and is performed concurrently with this procedure.

Therefore, the action of this step is only applicable when the figure is already being referenced. Fuel damaging , large amplitude power excursions may occur at low RPV pressures if reactor power is > 5%. Also , development of the HCTL assumes that the reactor is shutdown.

Therefore, depressurization due to HCTL or PSL during an ATWS is restricted to conditions where initial ATWS power level is 5%. The 5% determination is based on "initial ATWS power ," or reactor power before water level is lowered or boron is injected because there is no guarantee level will remain low or boron will remain in the core following depressurization.

Therefore, if initial ATWS power level is> 5%, depressurization is not required for this step. Control of suppression pool temperature is directed in the SPIT flowpath of E0-000-1 03. If the actions currently being taken in E0-000-1 03 to limit suppression pool temperature increase are inadequate and initial ATWS power is 5%, RPV pressure is reduced in order to remain below the HCTL.

LQ/P-5 E0-000-113 Revision 13 Page 37 of 67 If SRVs are used to maintain RPV pressure below the HCTL, the margin to the limit will not improve; that is, the delta between containment parameters and the HCTL will not increase. This is because SRV use adds heat to the suppression pool, and suppression pool temperature is one of the three factors that is included in the HCTL. Other methods of RPV depressurization that discharge outside primary containment, such as BPV's, and HPCI in CST to CST mode, reduce RPV pressure without adding heat to the suppression pool, thereby , increasing margin to the limit. The normal cooldown rate LCO may be exceeded to the extent necessary to maintain RPV pressure below the HCTL. If RPV pressure cannot be maintained below the HCTL, emergency RPV depressurization will be required, possibly resulting in an even more rapid cooldown. (

Reference:

PSTG 1st override before step RC/P-2) IF ALL:

• BORON INJECTION REQ'D

• MAIN CONDENSER AVAILABLE

• NO GROSS FUEL FAILURE

• NO MSL BREAK BYPASS MSIV AND CIG INTERLOCKS AND OPEN MSIV'S lAW ES-184-002 (ES-284-002)

This step is applicable to all subsequent steps within this flowpath.

It remains applicable to those steps until flowchart is exited. To stabilize and control RPV pressure, the reactor steam generation rate must remain within the capacity of systems designed to remove the steam from the RPV. With the reactor not subcritical, the amount of steam that may have to be released could be substantial.

If this heat energy is discharged to the suppression pool, the HCTL could be reached in a very short time. Therefore , utilization of the main condenser as a heat sink for this energy is of sufficient importance to warrant opening the MSIVs even if the valves automatically closed.

E0-000-113 Revision 13 Page 38 of 67 This step permits bypassing the low RPV water level portion of the MSIV isolation logic , and CIG isolation interlocks. OP-184-001 (OP-284-001

), Main Steam System, provides instructions for bypassing the following interlocks from the control room at 1 C644 (2C644) and 1 C645 (2C645):

• MSIV low water level isolations

• CIG high drywell pressure and low water level isolations.

ES-184-002 (ES-284-002), Reopening MSIVs Bypassing Isolations, provides instructions for restoring CIG (including a LOCA/LOOP compressor trip), and reopening the MSIVs. Other MSIV isolation interlocks are not bypassed because they provide automatic proteCtion for conditions where reopening the MSIVs is not appropriate. The determination of "NO GROSS FUEL FAILURE" or "NO MSL BREAK" can be made without long, detailed investigation.

Conservative implementation of this step would be to perform the procedure which bypasses interlocks in those cases that are not clearly determined, because re-opening MSIV's is very important during ATWS. MSIVs may be reopened if all the following conditions exist:

• Boron Injection is Required.

• Main condenser is available , since the only reason for opening the MSIVs is to utilize the main condenser as the heat sink.

• No indication of gross fuel failure, since opening the MSIVs with grossly failed fuel could result in a significant release of fission products to the environment. "Gross" fuel failure is specified to distinguish from small cladding leaks. The judgment is subjective based on operator assessment of all available indications. If it is concluded that no gross fuel failure exists but in actuality core damage has occurred, high radiation should be detected when the MSIVs are opened and MSIV isolation can be directed to reclose the valves.

LQ/P-6 E0-000-113 Revision 13 Page 39 of 67

• No indication of steam line break, since opening MSIVs with a break in the downstream pip i ng could result in an uncontrolled loss of reactor coolant inventory, release of fission products to the environment and cause personal injury or significant damage to plant equipment.

It is difficult, however, to determine whether a steam line break exists with the MSIVs closed, other than by visual inspection of system piping. lf.there is reasonable assurance that no break existed before MSIV closure, an operator may conclude that no break developed subsequent to valve closure. Still , the judgment is subjective.

based on assessment of all available indications.

If it is concluded that no steam line break exists but , in actuality , one does exist, high steam line flow and high steam tunnel temperature should be detected when MSIVs are opened and the MSIV isolation logic should automatically reclose the valves. (

Reference:

PSTG 2nd override before step RC/P-2) STABILIZE PRESS< 1087 PSIG USING BPV'S AUGMENTING PRESS CONTROL WITH ANY:

• SRV'S IF SUPP POOL LVL > 5' OPEN SRV'S USING OPENING SEQUENCE A B C IF 200 PSIG INDICATED ON EITHER ADS N2 BOTTLE HEADER PLACE ALL SRV SWITCHES TO OFF

• HPCI CST TO CST MODE

• RCIC CST TO CST MODE

• SJAE

• RFPT

• STEAM SEAL EVAP

• MSL DRAINS

• RWCU RECIRC MODE BYPASSING INTERLOCKS AS NECESSARY lAW ES-161-002 (ES-261-002)

• RWCU SLOWDOWN MODE ONLY WITH NO BORON INJECTED lAW ES-161-001 (ES-261-001)

CAUTIONS 50 ELEVATED SUPP CHMBR PRESS MAY TRIP RCIC ON HIGH EXHAUST PRESS.

E0-000-113 Revision 13 Page 40 of 67 The intent of this step is to hold pressure as constant as practicable.

RPV pressure is stabilized to facilitate control of RPV water level. If pressure is allowed to oscillate, RPV water level may fluctuate due to the effects of shrink and swell. Pressure is stabilized at a value below the high RPV pressure scram setpoint to avoid SRV actuation and to permit the scram logic to be reset (if no other scram signal exists). No minimum value is specified since the pressure at which RPV Control is entered cannot be predetermined and the procedure must provide appropriate guidance for all events. A target pressure should be selected close to the initial value and below the RPV pressure scram setpoint that permits use of available injection systems. An initial adjustment to establish an appropriate target pressure is permitted, p r ovided the target can be reached expeditiously and the Technical Specification cooldown rate LCO is not exceeded. A pressure reduction from an intermediate pressure to a pressure as low as the shutdown cooling RPV pressure interlock (98 psig) is not considered a permissible initial pressure adjustment even if the depressurization is within the LCO cooldown rate. To maintain RPV water level at intermediate or low RPV pressures, the typical complement of RPV injection systems should not require such a pressure reduction.

" Stabilize" means to hold RPV pressure as constant as practicable within due constraints imposed by the nature of the event, the degree of control afforded by the systems used, and the availability of personnel to perform manual control functions.

The specific actions required and the degree to which the ideal of a constant pressure can be approached will vary according to these constraints. Both the rate and the magnitude of RPV pressure changes must be considered. A pressure that is slowly decreasing over a relatively wide control band may be more " stable" than short-period oscillations within a narrower control band. In general , the adequacy of steps taken to stabilize RPV pressure must be judged by the effect of any continuing pressure variations on RPV water level and by whether additional actions are possible or likely to afford better control capability.

If pressure variations are not interfering with RPV water level or cannot be stopped, pressure may be considered stabilized. If continuing pressure oscillations are complicating efforts to control RPV water level or if the existing pressure prevents use of available injection systems, additional effort is warranted.

E0-000-113 Revision 13 Page 41 of 67 The main turbine bypass valves are the preferred means of controlling RPV pressure , since they provide good control capability , are of relatively large capacity, and do not add heat to the suppression pool. The direction to use the bypass valves implicitly permits opening the MSIVs and placing the main condenser in service if such actions are necessary and conditions permit. It does not, however, constitute authorization to defeat any MSIV isolation interlocks.

If the main turbine bypass valves cannot be used to control RPV pressure, or when the available capacity of the main turbine bypass valves (and main condenser) is less than that required to control RPV pressure below the high RPV pressure scram setpoint, additional systems must be employed to augment RPV pressure control. Since symptom-oriented procedures must accommodate a full spectrum of initial plant conditions and event scenarios, no prioritization regarding the use of the listed RPV pressure control systems is specified in this step. If suppression pool water level is not above the top of the SRV discharge device (5'), steam discharged through the SRVs passes directly into the suppression chamber airspace. The magnitude of the resultant primary containment pressure increase could potentially exceed primary containment pressure limits. When manual SRV actuation is required for RPV pressure control, an opening sequence is preferred which distributes heat uniformly throughout the suppression pool to avoid high local pool temperatures which may result in inefficient pool cooling. The opening sequence also uniformly distributes the total number of SRV actuations among the total number of Purpose of placing SRV switches to " OFF" when ADS nitrogen bottle supply drops to 200 psig is to preserve the remaining pneumatic supply in the 90 psig header so that the non-ADS SRVs will be available if RPV rapid depressurization . is later required and ADS valves do not work. The ADS nitrogen bottle supply is the safety-related backup to the 150-psig header. The bottles have a 3-day storage capacity based on the system design leakage rate. A drop in bottle pressure to 200 psig is indicative of a loss of system integrity and the potential loss of the bottles' ability to supply the ADS function. Manual and relief mode operation of SRVs uses pneumatic supply; safety" mode does not. By placing SRV control switches to " OFF," the SRVs will open at safety setpoints without expending pneumatic supply , thereby, preserving their availability to rapidly depressurize the RPV if later required.

If CIG compressors or lA system (if system crosstied) are in service and maintaining the 90 psig header, continued use of SRV's to stabilize pressure is allowed since the pneumatic supply for valve operation is maintained.

The SRV switch is returned to " OFF" when closed. This maintains depressurization function if header pressure is subsequently lost.

E0-000-113 Revision 13 Page 42 of 67 After the 90 psig CIG header is exhausted, use of the 150 psig ADS header is generally not authorized for pressure control. Instead, the 150 psig header is reserved for rapid depressurization.

The only exception to this is when shutting down from the remote shutdown panels. If pneumatic supply to non-ADS SRVs is lost, ON-1 00/200-009 directs operation of the ADS valves from the upper and lower relay rooms as a redundant path to achieve hot and cold shutdown for that event. No other procedure permits operating ADS SRVs from the relay rooms. For pressure control, HPCI and RCIC are operated in CST to CST mode. Operation of RWCU in the recirculation mode may require by-passing isolation interlocks.

In order to make this mode more effective cooling may need to be maximized.

Instructions for bypassing interlocks and maximizing cooling are contained in ES-161-002 (ES-261-002), RWCU Recirc Mode for Pressure Control. Since operation of RWCU in the recirculation mode does not remove coolant inventory from the RPV, overriding interlocks and operation of the system during boron injection (if the filter/demineralizers are bypassed) is authorized.

Operation of RWCU in the blowdown mode when boron has been injected into the RPV is not permitted in order to maintain the required concentration of boron in the RPV. Prior to RWCU blowdown, reactor coolant is sampled and analyzed for activity as prescribed by ES-161-001 (ES-261-001), RWCU Slowdown Mode Bypassing Interlocks.

Failure to determine coolant activity might result in discharge of radioactivity to the environment beyond allowable limits. This step does not require RPV cooldown nor does it provide all the necessary precautions that should be met prior to cooldown.

For guidance on RPV cooldown continue in this flowpath.

LQ/P-7 E0-000-113 Revision 13 Page 43 of 67 Instructions for operating the required systems are contained in the following procedures:

• BPVs OP-193-001 (OP-293-001)

Main Turbine Operation

• SRVs OP-183-001 (OP-283-001)

Automatic Depressurization System and Safety Relief Valves

• HPCI OP-152-001 (OP-252-001)

HPCI System

• RCIC OP-150-001 (OP-250-001)

RCIC System

• SJAE OP-172-001 (OP-272-001)

SJAE and Off Gas System

• Reactor Feed Pump Turbine OP-145-001 (OP-245-001)

RFP and RFP Lube Oil System

• Steam Seal Evaporator OP-192-00 1 ( OP-292-00 1 ) Seal Steam System

• Main Steam Line Drains OP-184-001 (OP-284-001)

Main Steam System (

Reference:

PSTG step RC/P-2) IF RX IS CRITICAL GO TO LQ/P-6 This step is applicable to all subsequent steps within this flowpath.

It remains applicable to those steps until flowchart is exited. The remaining steps of this flowpath depressurize and cool the RPV to cold shutdown conditions. The positive reactivity added during cooldown may return the reactor to criticality.

Stopping depressurization stabilizes RPV pressure until the reactor is again subcritical. (

Reference:

PSTG override preceding RC/P-3)

LQ/P-8 WHEN E0-000-113 Revision 13 Page 44 of 67 RX SUBCRITICAL DUE TO CONTROL RODS OR COLD S/D BORON WEIGHT INJECTED DEPRESSURIZE AT< 100°F/HR RPV depressurization and cooldown may not proceed until reactor is subcritical by control rod position or amount of boron injected into the RPV. If any amount of boron less than Cold Shutdown Boron Weight (CSBW) has been injected into the RPV, cooldown is not permitted.

CSBW is defined in LQ/Q-11 The core reactivity response from cooldown in a partially borated core is unpredictable and subsequent steps may not prescribe the correct actions for such conditions if criticality were to occur. If no boron has been injected into the RPV , depressurization and cooldown may proceed as long as the reactor is subcritical.

Such action is permitted even though the existing margin to criticality is small. A return to criticality under these conditions is acceptable because termination of the cooldown will stop the reactor power increase. If SRV handswitches had been placed to " OFF" position per LQ/P-6 , the requirement to maintain switches in " OFF" no longer applies. Handswitches are placed to "OFF" per LQ/P-6 to 'preserve the remaining pneumatic supply during the stabilization step so SRVs would be available if later required for rapid depressurization. (See LQ/P-6 basis.) When RPV parameters have been stabilized and it has been decided to initiate a depressurization , the use of SRVs to depressurize would be consistent with the purpose they were preserved to perform, i.e., depressurize the RPV. The pressure reduction performed in this step may be carried out using any of the systems listed in step LQ/P-6. The technical specification cooldown rate is observed to maintain RPV metal ductility limits. (

Reference:

PSTG RC/P-3)

LQ/P-9 LQ/P-10 IF SRV'S ARE BEING USED TO DEPRESSURIZE USE SUSTAINED SRV OPENING E0-000-113 Revision 13 Page 45 of 67 Purpose of maintaining an SRV open when using SRVs to depressurize during the established cooldown is to conserve the remaining pneumatic supply. This prolongs the period of time that the Operator can control pressure from the control room while attempts are made to recover CIG. Maintaining an SRV open also completes the depressurization, or a major portion of it, while pneumatic supply is available rather than delaying depressurization and risking the ability to do so. Sustained SRV opening also minimizes the number of challenges to SRV operability.

This step recognizes that SRVs might be in use as directed by other steps in this procedure.

It does not give authorization to exceed the 1 00°F/hr cooldown limit. Steps that authorize SRV use specify whether the 1 00°F/hr cool down rate may be exceeded or not. (

Reference:

PSTG step RC/P-3) WHEN SID COOLING INTERLOCK CLEARS CONTINUE TO COOL DOWN TO < 200°F WITH S/D COOLING USING ONLY THOSE RHR PUMPS NOT REQ'D TO MAINTAIN LVL > +13" IN LPCI MODE Operation of shutdown cooling is the normal method of conducting a controlled cooldown of the RPV to cold shutdown conditions. Shutdown cooling is placed in service when the low RPV water level and high RPV pressure shutdown cooling interlocks clear. Maintaining RPV water level in the preferred band takes priority over forced cooldown of the RPV. If operation of an RHR pump in the LPCI Mode is required to maintain water level, it may not be used for shutdown cooling. Loss of Shutdown Cooling at +13" is a re-entry condition to E0-1 00-113. Instructions for placing shutdown cooling in service are found in OP-149-002 (OP-249-002), RHR Shutdown Cooling. (

Reference:

PSTG step RC/P-4)

LQ/P-11 IF SID COOLING CANNOT BE PLACED IN SERVICE MAINTAIN PRESS< SID COOLING INTERLOCK WITH SYSTEMS USED FOR DEPRESS E0-000-113 Revision 13 Page 46 of 67 If shutdown cooling cannot be established, continued RPV depressurization and cooldown may be accomplished using any combination of the systems listed in Step LQ/P-6. The wording of this step permits, but does not require , continued cooldown below the shutdown cooling RPV pressure interlock.

If no requirement for rapid depressurization has been established , continued pressure reduction is expected to be of little benefit. As RPV pressure and temperature decrease , it may be necessary to re-evaluate the most appropriate method for continu i ng to achieve further pressure and temperature reduction.

Under certa i n conditions RPV pressure may be reduced to zero and temperature to below 212°F without employing shutdown cooling. (

Reference:

PSTG step RC/P-4)

CONTROL ROD INSERTION 113 SH 2 CR-1 IS MORE THAN 1 CONTROL ROD> 00 E0-000-113 Revision 13 Page 47 of 67 This step becomes effective when read and remains effective until flowchart is exited. The intent of this step is to define the condition that allows E0-000-113 sheet 2 to be exited. Refer to step LQ-2. (

Reference:

PSTG first override before step RC/Q-1) CR-2 STOP BORON INJECTION AND GO TO SCRAM PROCEDURE This procedure may be used concurrently with boron injection steps. When this procedure is successful in inserting rods then it is appropriate to terminate boron injection.

This step directs entry to the scram procedure where guidance will be given for balance of plant systems. The scram procedure was exited when it was determined that the reactor would not remain shut down under all conditions without boron and can now be returned to and performed concurrently with other emergency procedures. (Reference

PSTG first override before step RC/Q-1) CR-3 USE ONE OR MORE OF FOLLOWING The options available for control rod insertion are listed and conditions must be evaluated to determine which method or methods will be most effective. (

Reference:

PSTG step RC/Q-7)

E0-000-113 Revision 13 Page 48 of 67 DRIFTING RODS WITH COOLING WATER PRESS CR-4 MAXIMIZE CRD USING EITHER UNIT This step is provided to increase cooling water differential pressure to maximum which may cause the control rods to drift inward and shutdown the reactor. Instructions for maximizing CRD flow are given in the operating procedure OP-155-001 (OP-255-001

), Control Rod Drive Hydraulic System. (

Reference:

PSTG step RC/Q-7.2)

CR-5 IF DRIVING RODS CRD RUNNING IN EITHER UNIT DRIVE RODS AS FOLLOWS E0-000-113 Revision 13 Page 49 of 67 This step tests the availability of CRD. CRD pumps must be running to provide the drive water differential pressure necessary to insert control rods. If CRD is not running, the remaining steps need not be performed.

Instructions for cross-tying Unit 1 and Unit 2 CRD Systems are given in OP-155/255-001. (

Reference:

PSTG step RC/Q-7.2) CR-6 BYPASS RWM (HC) Rod Worth Minimizer must be bypassed to remove control rod blocks that are generated from the abnormal rod pattern. RWM is bypassed using the installed bypass key switch lAW OP-131-001 (OP-231-001), Rod Worth Minimizer. (

Reference:

PSTG step RC/Q-7.2)

CR-7 ESTABLISH APPROXIMATELY (IF OBTAINABLE)

(HC) 63 GPM COOLING WATER FLOW AND 350 PSID DRIVE WATER PRESS CLOSING CHARGING WATER ISO 146F034 (246F034)

AS NECESSARY Intent of this step is to ensure CRD system flow and differential pressure is sufficient to manually drive rods. System Operating Parameters are listed preceded by " approximately" to recognize that the system may not be operating within these limits when this step is reached. Such is the case if CRD flow was maximized to drift rods. If rods can be manually inserted, CRD parameters are within acceptable limits. Drive water pressure of 350 psid is directed due to the potential for channel bow/bulge.

There may be difficulty inserting rods manually at 250 psid due to this phenomenon.

Increasing drive water pressure is already directed in ON-155-001 (ON-255-001) if difficulty is encountered inserting rods. The higher drive water pressure potentially reduces plant risk by minimizing the time to drive rods.

E0-000-113 Revision 13 Page 50 of 67 In order to manually drive control rods, charging water flow must be isolated from the HCUs so cooling water flow and drive water pressure can be restored.

If ATWS is electrical or scram has been reset, scram valves will not be open. If a scram signal is in, scram valves will be open with charging water flowing through the CRDs into the RPV. Even if scram valves to only one CRD are open, it may not be possible to establish normal drive water pressure.

For this reason it may be necessary to close the charging water isolation , 146F034 (246F034) valve. Status of scram valves can be determined selecting DISPLAY SCRAM VALVES OPEN. (Reference

PSTG step RC/Q-7.2) , CR-8 SELECT RODS IN ROTATING QUADRANTS (HC) CR-9 CR-10 AND DEPRESS CONTINUOUS INSERT FOR EACH OF FOLLOWING UNTIL FULL-IN OR ROD WILL NOT MOVE: 1 INTERMEDIATE POSITION RODS 2 FULL-OUT RODS A sequence for control rod insertion is specified that minimizes flux imbalance while control rods are being inserted.

Regardless of intermediate position , all rods with intermediate indication on the full-core display (not red or green) are inserted first followed by control rods that have (red) full out indication, in a symmetrical pattern. This is because intermediate rods typically have a higher rod worth than full-out rods. (

Reference:

PSTG step RC/Q-7.2)

IS ADEQUATE CRD PUMP SUCTION AVAILABLE Normal CRD suction will be lost when CST level drops below 45% (135,000 gallons).

CRD suction is preserved by making up to the hotwell with demin water transfer and jockey pumps. (

Reference:

PSTG step RC/Q-7.2)

ENSURE AT LEAST ONE COND PUMP RUNNING The intent of this and the next two steps is to maintain CRD suction for driving rods. Condensate Pump provides reject flow to the reject line. (

Reference:

PSTG step RC/Q-7.2)

CR-11 CR-12 CR-13 OPEN HV10517 (HV20517)

HP CDSR DEMIN WATER MAKEUP E0-000-113 Revision 13 Page 51 of 67 Opening this valve opens a flow path from the demineralized water storage tank to the hotwell. (

Reference:

PSTG step RC/Q-7.2)

START ALL:

• OP507 A DEMIN WTR TRANSFER PP

• OP507B DEMIN WTR TRANSFER PP

• OP508 DEMIN WTR JOCKEY PP Operation of the demin water transfer and jockey pumps provides makeup water to the hotwell. (

Reference:

PSTG step RC/Q-7.2)

WHEN RODS WILL NOT INSERT (HC) ENSURE CHARGING WATER ISO 146F034 (246F034)

OPEN When this method of control rod insertion is not needed or is not effective, restore the charging water header flow by opening charging water header isolation valve 146F034 (246F034 ). (

Reference:

PSTG step RC/Q-7.2)

INDIVIDUALLY SCRAMMING RODS CAUTION SDV VENT AND DRAIN VALVES FAIL CLOSED ON LOSS OF 1-A. E0-000-113 Revision 13 Page 52 of 67 Auxiliary bus load shed causes loss of instrument air. Scram discharge vent and drain valves fail closed upon loss of instrument air , thereby preventing the volume from draining.

Inability to drain the scram discharge volume will prevent two Control Rod Insertion options: Individually Scramming Rods; Resetting and Scramming Again. Purpose of this caution is to advise the operator of the nega t ive impact on control rod insertion from failure to promptly recover from the loss of instrument air. It is recognized that instrument air can be lost for many reasons, but in Level/Power Control, the probability of losing instrument air is increased since level is intentionally lowered close to -129" and containment pressure is likely increasing close to 1. 72 psig. (

Reference:

PSTG Caution 1 0) CR-14 DISABLE ARI CR-15 AND BYPASS RPS LOGIC TRIPS AS NECESSARY lAW ES-158-002 (ES-258-002)

If a scram signal exists, it must be reset before individually scramming rods. The presence of ARI and RPS logic trips will prevent a SCRAM signal from being reset. ES-158-002 (ES-258-002), RPS and ARI Trip Bypass, provides instructions to disable ARI and bypass RPS logic trips. (Reference

PSTG step RC/Q-7.2)

RESET SCRAM After ARI is disabled and RPS logic trips are bypassed , the scram is reset. Instructions for resetting the scram are given in OP-158-001 (OP-258-001

), RPS System. (

Reference:

PSTG step RC/Q-7.2)

CR-16 IF ALL:

• ARI DISABLED

• SCRAM RESET

• SCRAM AIR HEADER NOT VENTED

• RPS FUSES NOT REMOVED INDIVIDUALLY SCRAM RODS lAW ATTACHMENT A E0-000-113 Rev i s i on 13 Page 53 of 67 The advantage of individually scramming control rods is that resulting exhaust water occupies a smaller volume than required for a full scram. While it may not be possible to scram all control rods in a full scram, it may be possible to scram some rods individually.

Individual control rod scrams using the scram test switches will generally insert control rods faster than manual control rod insertion.

Even with an incompletely drained or non-vented scram discharge volume, sufficient capacity may remain in the scram discharge volume to accommodate discharge flow from control rod scrams taken one at a time as opposed to all at once. Attachment A describes one sequence for manually scramming control rods. This sequence insures that control rods are inserted in a symmetrical pattern that balances the flux in each quadrant of the core. It is important to attempt to scram each contro l rod in a particular group designation (A,B,C, or D) prior to attempting to scram control rods for a different group designation even if some control rods will not insert. Control rods are individually scrammed until control rods will not insert and it is determined that other methods of inserting control rods would be more effective. (

Reference:

PSTG step RC/Q-7.2)

RESETTING AND SCRAMMING AGAIN CAUTION SDV VENT AND DRAIN VALVES FAIL CLOSED ON LOSS OF 1-A. E0-000-113 Revision 13 Page 54 of 67 Auxiliary bus load shed causes loss of instrument air. Scram discharge vent and drain valves fail closed upon loss of instrument air, thereby preventing the volume from draining.

Inability to drain the scram discharge volume will prevent two Control Rod Insertion options: Individually Scramming Rods; Resetting and Scramming Again. Purpose of this caution is to advise the operator of the negative impact on control rod insertion from failure to promptly recover from the loss of instrument air. It is recognized that instrument air can be lost for many reasons, but in Level/Power Control, the probability of losing instrument air is increased since level is intentionally lowered close to -129" and containment pressure is likely increasing close to 1.72 psig. (

Reference:

PSTG Caution 1 0) CR-17 WHEN SCRAM FAILURE IS HYDRAULIC CONTINUE The following steps attempt to repeat a manual scram. They are only applicable for a hydraulic scram failure. Table 20 is provided to help determine if the ATWS is electrical or hydraulic.

Check the answer to each line according to plant conditions and the column that has the most checks is the most likely ATWS type. TABLE 20 SYMPTOMS OF ELECTRICAL OR HYDRAULIC ATWS CONDITION

• FULL CORE DISPLAY FULL-IN/FULL-OUT MODE

• SCRAM AIR HEADER LOW PRESS ALARM

• SDV VENT AND DRAIN VLV'S CLOSED

• ALL ARI VL V'S REPOSITIONED (

Reference:

PSTG step RC/Q-7.2)

HYD YES YES YES YES ELEC NO NO NO NO CR-18 CR-19 CR-20 CR-21 DISABLE ARI AND BYPASS RPS LOGIC TRIPS AS NECESSARY lAW ES-158-002 (ES-258-002)

E0-000-113 Revision 13 Page 55 of 67 If a scram signal exists, it must be reset before attempting to scram again. The presence of ARI and RPS logic trips will prevent a SCRAM signal from being reset. ES-158-002 (ES-258-002), RPS and ARI Trip Bypass , provides instructions to disable ARI and bypass RPS logic trips. (

Reference:

PSTG step RC/Q-7.2) RESET SCRAM After ARI is disabled and RPS logic trips are bypassed, the scram is reset. Instructions for resetting the scram are given in OP-158-001 (OP-258-001

), RPS System. (Reference

PSTG step RC/Q-7.2)

ENSURE CHARGING WATER ISO 146F034 (246F034)

OPEN If the charging water isolation valve was closed for driving rods, the valve will need to be opened to charge the HCU accumulators. WHEN SDV PARTIALLY DRAINS INSERT MANUAL SCRAM Since it is possible that a potential hydraulic problem exists , the scram discharge volume is allowed to drain and then a manual scram is inserted again to attempt to insert the control rods. How long to wait for SDV to partially drain after the scram is reset can be estimated based on the following:

• Approximately 45 seconds are required for the scram air header to fully repressurize.

• Approximately 5 minutes are required for HCU accumulators to fully recharge.

• When the control room alarm "RPS CHAN A1/A2 (B1/B2) SCRAM DSCH VOL HI WTR LEVEL TRIP" is clear then less than 66 gallons remain in the SDV. This normally takes 7 minutes to occur. (

Reference:

PSTG step RC/Q-7.2)

CR-22 VENTING SCRAM AIR HEADER E0-000-113 Revision 13 Page 56 of 67 Steps to vent and restore Scram Air Header are placed in upper left corner of flowchart, upside down, for use by the PCO, as necessary, without having to turn board around. WHEN SCRAM FAILURE IS ELECTRICAL CONTINUE The following steps attempt to vent the scram air header. They are only applicable when the scram air header is not already vented by automatic systems as would be the case when the ATWS is due to electrical systems failure. Table 20 is provided to help determine if the ATWS is electrical or hydraulic.

Check the answer to each line according to plant conditions and the column that has the most checks is the most likely ATWS type. TABLE 20 SYMPTOMS OF ELECTRICAL OR HYDRAULIC ATWS CONDITION

• FULL CORE DISPLAY FULL-IN/FULL-OUT MODE

• SCRAM AIR HEADER LOW PRESS ALARM

• SDV VENT AND DRAIN VLV'S CLOSED

• ALL ARI VL V'S REPOSITIONED (

Reference:

PSTG RC/Q-7.2)

HYD YES YES YES YES ELEC NO NO NO NO (1) CR-23 (1) CR-24 C) CR-25 E0-000-113 Revision 13 Page 57 of 67 OPEN ARI SYS SOLENOID VALVES BYPASS 147021 (247021) (HC} Initiation of ARI opens both SCRAM Air Hdr Vent valves SV-147099/SV-147100 (SV-247099/SV-247100) and closes both SCRAM Hdr Block Valves SV-147101/SV-147102 (SV-247101/SV-247102).

Although unlikely, ARI failure could be due to one or both vent valves failing to open while both block valves properly close. If both ARI block valves are closed , a flowpath does not exist for venting the SCRAM air header. Due to this case, the block valves' bypass is opened before proceeding. (

Reference:

PSTG/SAG RC/Q-7.2)

CLOSE SCRAM AIR SUPPLY (HC} 147002A (247002A)

AND 147002B (247002B)

AND UNCAP AND OPEN SCRAM AIR HEADER VENT 147007 (247007) This step attempts to open scram valves by removing air from the scram air header using manual valves. (Reference

PSTG RC/Q-7.2)

WHEN RODS STOP MOVING INWARD (HC} 1 CLOSE AND CAP SCRAM AIR HEADER VENT 147007 (247007) 2 OPEN SCRAM AIR SUPPLY 147002A (247002A)

OR 147002B (247002B) 3 CLOSE ARI SYS SOLENOID VALVES BYPASS 147021 (247021) This step is simply to restore from actions taken in previous step, when these actions are no longer warranted as indicated by control rods not moving inward whether they did move or not. (

Reference:

PSTG RC/Q-7.2)

CR-26 CR-27 WHEN DE-ENERGIZING SCRAM SOLENOIDS SCRAM FAILURE IS ELECTRICAL REMOVE RPS FUSES lAW ES-158-001 (ES-258-001) E0-000-113 Revision 13 Page 58 of 67 This step attempts to open scram valves by de-energizing the scram pilot solenoids. It is only applicable when the scram air header is not already vented by automatic systems as would be the case when the A TWS was due to electrical systems failure. Direction for removing the fuses to de-energize the scram pilot solenoids is given in ES-158-001 (ES-258-001

), De-energizing Scram Pilot Solenoids. Table 20 is provided to help determine if the ATWS is electrical or hydraulic.

Check the answer to each line according to plant conditions and the column that has the most checks is the most likely ATWS type. TABLE 20 SYMPTOMS OF ELECTRICAL OR HYDRAULIC ATWS CONDITION

• FULL CORE DISPLAY FULL-IN/FULL-OUT MODE

• SCRAM AIR HEADER LOW PRESS ALARM

• SDV VENT AND DRAIN VALVES CLOSED

• ALL ARI VALVES REPOSITIONED (Reference

PSTG step RC/Q-7.2)

WHEN RODS STOP MOVING INWARD RESTORE RPS FUSES lAW ES-158-001 (ES-258-001)

HYD YES YES YES YES ELEC NO NO NO NO This step is simply to restore from actions taken in previous step, when these actions are no longer warranted as indicated by control rods not moving inward whether they did move or not. (

Reference:

PSTG step RC/Q-7.2)

CR-28 WHEN VENTING CRD OVERPISTON VOLUME ALL OTHER METHODS HAVE FAILED INSERT CONTROL RODS lAW ES-155-001 (ES-255-001)

E0-000-113 Revision 13 Page 59 of 67 This step is entered when other means of inserting control rods have been unsuccessful.

This procedure provides a method of inserting the control rods using the differential pressure available between the reactor and the reactor building atmosphere.

By connecting a hose to the withdrawal line vent and directing the effluent to a contained radwaste drain , the reactor pressure is used to insert the control rod. This method is used only when all other methods for inserting control rods are unsuccessful because it is difficult , time consuming , dangerous or could cause the spread of contamination.

Instructions for venting CRD overpiston volume are found in ES-155-001 (ES-255-001

), Venting CRD to Insert Control Rods. (Reference

PSTG step RC/Q-7.2)

E0-000-113 Revision 13 Page 60 of 67 4. REFERENCES 4.1 EC-EOPC-0519 , SABRE Calculations to Support Technical Basis of IPE and ATWS EOP 4.2 EC-RISK-1 063, Evaluation of Operator Actions for Application in the Susquehanna Individual Plant Examination 4.3 LRF 99-055-034

This LRF generated two placards that include instructions to vent the Unit 1 and Unit 2 SCRAM air headers, respect i vely. Placards are posted at the Unit 1 and Unit 2 CRD skids. 4.4 Memo PLI-50399 , C. Kukielka to B.R. Stitt , "Operation of Diesel Generators Without ESW Flow ," May 13, 1987 D INDIVIDUALLY SCRAM RODS CAUTION Att ac hment A E 0-000-113 Revision 13 Page 61 of 67 Attempt to scram all control rods for any group designation (A , B , C, or D) before attempting to scram any control rods from a different group. ..----26-51 22-47 D B 30-47 30-55 18-43 A c c 26-55 22-39 14-51 A D D 22-51 26-39 18-47 c A D 26-47 22-43 14-43 c A B 26-43 30-31 18-35 B A A 30-39 26-35 22-35 c D c 30-23 26-23 26-31 c A c 26-27 22-27 18-27 B A c 22-19 18-19 22-23 c A D 26-15 14-19 18-23 c D B 22-11 22-15 10-19 A B B 26-07 30-07 18-15 A c D 26-19 26-11 14-11 D B B 30-15 18-11 A c 22-07 D FOR EACH HCU: 18-51 10-51 A B 22-55 14-47 D c 18-55 10-47 B A 10-43 06-43 D c 14-39 06-47 A D 14-35 10-39 D c 18-39 06-35 B A 22-31 10-35 B B 14-27 18-31 B D 14-23 14-31 A c 10-27 02-27 D A 10-23 06-23 c B 06-27 02-19 c c 10-1 1 06-19 D A 14-15 10-15 c A 18-07 14-07 B A 14-55 A 26-59 B 18-59 c 22-59 A 02-43 A 02-39 D 02-35 c 06-39 B 06-31 D 02-31 B 10-31 A 02-23 D 18-03 A 22-03 c 06-15 D 26-03 D CRD MAP UNIT 1 AND UNIT 2 NORTH=> REACT O R VESS E L 30-59 D 34-59 A 38-59 c 54-43 A 42-59 D 54-31 B 50-35 A 54-39 D 58-43 B 58-35 D 54-27 A 58-27 B 58-19 D 42-03 B 34-03 c 50-11 c 0 1. PLACE BOTH SCRAM TEST SWITCHES TO TEST. 0 2. WAIT 10 SECONDS. 46-55 A 42-51 B 50-51 A 54-47 B 50-43 c 50-39 B 46-35 D 50-31 D 54-35 c 58-31 c 58-23 B 54-19 c 38-03 A 54-15 B 46-11 B 30-03 B 0 3. RETURN BOTH SCRAM TEST SWITCHES TO NORMAL. Page 1 of 1 r--38-55 B 38-51 34-51 A c 42-55 42-47 c A 46-51 46-47 D c 46-43 42-43 B D 50-47 46-39 D A 46-31 38-35 c A 42-35 42-31 B A 58-39 38-27 A c 54-23 42-23 D c 50-27 38-19 c A 50-19 46-27 A B 50-15 50-23 D B 46-19 42-1 9 D B 46-07 46-15 A c 42-11 42-07 D c 34-55 30-51 D B 38-47 34-47 D B 38-43 34-43 c A 42-39 38-39 c B 38-31 30-43 D D 34-35 34-39 c D 34-31 30-35 B B 38-23 30-27 B D 34-27 34-23 A D 42-15 30-19 A B 46-23 34-19 A c 42-27 38-15 D D 38-11 34-15 c B 38-07 30-11 B D 34-11 34-07 A D

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