ML102510161

Request for Operating License Amendment - Revision of Feedwater Line Break with Loss of Offsite Power and Single Failure Analysis

ML102510161
Person / Time
Site:	Palo Verde
Issue date:	08/27/2010
From:	Mims D Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-06244-DCM/RAB/DFS
Download: ML102510161 (45)

Text

10 CFR 50.90 AA M A subsidiaryof PinnacleWest CapitalCorporation Dwight C. Mims Mail Station 7605 Palo Verde Nuclear Vice President Tel. 623-393-5403 P.O. Box 52034 Generating Station Regulatory Affairs and Plant Improvement Fax 623-393-6077 Phoenix, Arizona 85072-2034 102-06244-DCM/RAB/DFS August 27, 2010 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529, and 50-530 Request for Operating License Amendment - Revision of Feedwater Line Break with Loss of Offsite Power and Single Failure Analysis Pursuant to 10 CFR 50.90, Arizona Public Service Company (APS) hereby requests to amend Operating Licenses NPF-41, NPF-51, and NPF-74 for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, respectively. The proposed amendment incorporates a revision to an element of the methodology in the feedwater line break with loss of offsite power and single failure (FWLB/LOP/SF) event analysis summarized in the PVNGS updated final safety analysis report (UFSAR) Chapter 15, Section 15.2.8.

The proposed amendment would revise the FWLB/LOP/SF analysis by assuming operator action during the event at 20 minutes to control pressurizer level. The previous analysis assumed operator action at 30 minutes. The proposed change is the result of a lack of a design basis analyzed value to use for the reactor coolant system (RCS) bleed-off rate assumed in the FWLB/LOP/SF event analysis, which has resulted in APS conservatively setting the bleed-off rate at zero gallons per minute. The use of that

'assumption in the FWLB/LOP/SF analysis could challenge the continued operability of the pressurizer safety valves due to overfilling of the pressurizer. Operator action at 20 minutes will ensure adequate control of the RCS heat-up and pressurizer level. The proposed change involves a "departure from a method of evaluation described in the UFSAR," as defined in 10 CFR 50.59(a)(2). Therefore, pursuant to 10 CFR 50.59(c)(2)(viii), the proposed change requires NRC approval, by license amendment, prior to implementation.

APS has evaluated continued operation and determined that the condition addressed by this proposed amendment does not pose an operability concern. This is based on the existing flowpath from the reactor coolant pump bleed-off line to the reactor drain tank, A member of the STARS (strategic Teaming and Resource Sharing) Alliance AoAa * ,, /

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ATTN: Document Control Desk U.S. Nuclear Regulatory Commission

Subject:

Request for Operating License Amendment - Revision of Feedwater Line Break with Loss of Offsite Power and Single Failure Analysis Page 2 not credited in the FWLB/LOP/SF analysis, that will reduce the potential rate of level increase in the pressurizer. In addition, the current PVNGS emergency operating procedures (EOPs) contain explicit directions to the operators to ensure that the plant is placed in a stable, safe condition following a feedwater line break (FWLB) event. The EOP directions are parameter dependent and not time dependent. These directions are not affected by the proposed amendment and they ensure that the required safety functions continue to be met.

Approval of the proposed amendment is requested by August 28, 2011. Once approved, the amendment shall be implemented within 90 days.

In'accordance with the PVNGS Quality Assurance Program, the Plant Review Board and the Offsite Safety Review Committee have reviewed and concurred with this proposed amendment. By copy of this letter, this submittal is being forwarded to the Arizona Radiation Regulatory Agency (ARRA) pursuant to 10 CFR 50.91 (b)(1).

The following commitment is being made to the NRC in this letter:

Upon NRC approval, the operator action completion time will be added to the Palo Verde Time Critical Action Program.

Should you need further information regarding this amendment request, please contact Russell Stroud, Licensing Section Leader, at (623) 393-5111.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on __________ -

(Date) "

Sincerely, DCM/RAS/DFS/

Enclosure:

Evaluation of the Proposed Change cc: E. E. Collins Jr. NRC Region IV Regional Administrator J. R. Hall NRC NRR Senior Project Manager L. K. Gibson NRC NRR Project Manager J. H. Bashore NRC Senior Resident Inspector for PVNGS A. V. Godwin Arizona Radiation Regulatory Agency (ARRA)

T. Morales Arizona Radiation Regulatory Agency (ARRA)

ENCLOSURE Evaluation of the Proposed Change

Subject:

Request for Operating License Amendment to Updated Final Safety Analysis Report (UFSAR) Chapter 15 Feedwater Line Break Analysis with Loss of Offsite Power and Single Failure (FWLB/LOP/SF) Analysis 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 - Conclusions

6.0 REFERENCES

ATTACHMENT:

UFSAR Chapter 15 Page Markups

Enclosure Evaluation of the Proposed Change 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating Licenses NPF-41, NPF-51, and NPF-74, for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, respectively.

The proposed amendment incorporates a revision to an element of the methodology in the feedwater line break with loss of offsite power and single failure (FWLB/LOP/SF) event analysis summarized in the updated final safety analysis report (UFSAR) Chapter 15, Section 15.2.8. The proposed amendment would revise the FWLB/LOP/SF analysis by assuming operator action during the event at 20 minutes to control pressurizer level.

The previous analysis assumed operator action at 30 minutes.

The proposed change is the result of a lack of a design basis analyzed value to use for the reactor coolant system (RCS) bleed-off rate assumed in the FWLB/LOP/SF event analysis, which has resulted in Arizona Public Service Company (APS) conservatively setting the bleed-off rate at zero gallons per minute>- The use of that assumption in the FWLB/LOP/SF analysis could challenge the continued operability of the pressurizer safety valves (PSVs) due to overfilling of the pressurizer. Operator action at 20 minutes will ensure adequate control of the RCS heat-up and pressurizer level.

2.0 DETAILED DESCRIPTION The proposed license amendment would revise UFSAR Chapter 15, Section 15.2.8 to incorporate a revision to an element of the methodology for the FWLB/LOP/SF analysis by assuming operator action during the event at 20 minutes to control pressurizer level.

The previous analysis assumed operator action at 30 minutes.

In 2005, a question was raised by APS about the assumptions made in the UFSAR Chapter 15, Section 15.5.2.3.B analysis concerning control of reactor coolant pump (RCP) bleed-off rates. For that specific concern, the result of that evaluation determined the control of RCP bleed-off rates was acceptable as provided. However, as part of the extent of condition evaluation of that question, the FWLB/LOP/SF event, reported in UFSAR Chapter 15.2.8.4, was determined to require evaluation with respect to its RCP seal bleed-off assumptions. No other events were identified as being impacted by this RCP seal bleed-off question.

The FWLB/LOP/SF event is the limiting decrease in heat removal event that may challenge the PSVs operability and RCS pressure boundary integrity due to over filling of the pressurizer. As a result of this evaluation, it was determined that the assumption used for RCP bleed-off rate was less than sufficiently conservative. A lack of a design basis analyzed value to use for the system bleed-off rate has resulted in APS conservatively setting the bleed-off rate at zero gallons per minute. The use of that assumption in the FWLB/LOP/SF analysis could challenge the continued operability of the PSVs due to overfilling of the pressurizer.

The apparent cause of the RCS bleed-off rate assumed in the FWLB/LOP/SF analysis being less than sufficiently conservative is related to changing the computer code used from CESEC (Reference 6.1) to CENTS (Reference 6.2). In the NRC letter, "Palo 2

Enclosure Evaluation of the Proposed Change Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments, RE:

Various Administrative Controls," dated October 15, 2001 (ADAMS Accession No. ML012880473) (Reference 6.3), license amendment 137 Was approved, which added the CENTS computer code as an acceptable code methodology. However, there was a variance in how the two codes deal with some inputs and assumptions, which was not identified by APS prior to use of the CENTS code. In the CESEC code, the RCP seal bleed-off is part of the letdown flow assumption and as such, was considered isolated when letdown was isolated. In the modeling of FWLB/LOP/SF, letdown is always isolated as part of the LOP assumption. In the CENTS computer code, it was not realized that the code contained a separate variable that did not isolate RCP seal bleed-off with letdown in this event. This resulted in an assumed RCP seal bleed-off rate being credited in the analysis which is not conservative.

APS has reviewed the various PVNGS analyses that are performed with the CENTS code to ensure that this modeling assumption is not credited, and to ensure that there are no other elements of the CENTS code that would introduce other non-conservative assumptions. No similar assumptions were identified in that review. In addition, procedural precautions have been added to the design process to require enhanced review of any code related changes. These precautions specifically highlight verifying that the code does not result in non-conservative use of any assumptions or inputs to the code.

To resolve the FWLB/LOP/SF analysis non-conservative assumption, the analysis was revised and resulted in crediting operator action at 20 minutes to control RCS heat-up and pressurizer level increases. The 20 minutes replaces the 30 minutes previously credited. The operator action at 20 minutes ensures that the pressurizer level does not rise to a level where water would be passed through the PSVs, which would result in the PSVs being inoperable. The change in the assumed operator action time involves a "departure from a method of evaluation described in the UFSAR," as defined in 10 CFR 50.59(a)(2). Therefore, pursuant to 10 CFR 50.59(c)(2)(viii), the change requires NRC approval, by license amendment, prior to implementation.

APS has evaluated continued operation and determined that the condition addressed by this proposed amendment does not pose an operability concern. This is based on an existing flowpath from RCP bleed-off line to the reactor drain tank (RDT), not credited in the FWLB/LOP/SF analysis, that will reduce the potential rate of increase in the pressurized level. In addition, the current PVNGS emergency operating procedures (EOPs) contain explicit directions to the reactor operators to ensure that the plant is placed in a stable, safe condition following a feedwater line break (FWLB) event. That direction includes actions to control RCS heat-up and maintain the pressurizer level below a level that could pass water through the primary safety valves and potentially challenge the RCS inventory control. These EOP actions are parameter dependent and not time dependent. The proposed change does not affect the EOP directions for continued control of RCS heat-up rate and pressurizer level increase during the FWLB event, and ensures that all required safety functions continue to be met.

3

Enclosure Evaluation of the Proposed Change

=

System Description===

The FWLB/LOP/SF event, discussed in UFSAR Chapter 15.2.8.4, assumes a RCP seal bleed-off flow of approximately three gallons per minute throughout the event. That analysis also assumes that a containment isolation actuation signal (CIAS), safety injection actuation signal (SIAS) and main steam isolation signal (MSIS) occurs at the time of a reactor trip and a LOP occurs three seconds later (early in the event).

In UFSAR Section 9.3.4.2.3.2.G, it states that the RCP controlled bleed-off (CBO) header relief valve located on the RCP CBO header redirects flow to the RDT in the event that the normal flowpath to the volume control tank (VCT) is isolated. It states that this relief valve serves no overpressure protection function and that it is sized to pass the flow rate from the failure of two seal stages in one RCP plus the normal bleed-off from the other three RCPs. The relief valve set pressure is greater than the normal operating pressure of the header (aligned to the VCT) and less than the CBO high-high pressure alarm. However, even though this flow path will pass some of the RCP seal bleed-off flowrate, the flowrate through that relief valve during those conditions has not been quantified (i.e., no design basis or engineering calculation, etc). As a result, a zero RCP seal bleed-off flowrate should have been assumed in the FWLB/LOP/SF analysis. To resolve this issue the FWLB/LOP/SF analysis was revised as discussed below.

The Analysis UFSAR Section 15.2.8.4 provides the acceptance criteria for the FWLB/LOP/SF event which requires (without operator action) that the PSVs do not pass water and the pressurizer does not become solid for a minimum of 30 minutes. In the FWLB/LOP/SF analysis,'the event is conservatively modeled with adverse assumptions for many systems, structures, and components (SSCs) and no credit is taken for operator action for the first 30 minutes of the event [i.e., opening the atmospheric dump valves (ADVs),

securing charging flow, etc.]. In the FWLB!LOP/SF analysis, the event is terminated after 30 minutes (1800 seconds) when operators initiate a controlled cool down to shutdown cooling entry conditions. The analysis shows that during the event the reduction of primary side heat removal causes RCS volumetric expansion that results in compression of the pressurizer steam volume due to insurge flow through the pressurizer surge line. This increases the pressurizer pressure above the PSV opening setpoints and causes the PSVs to open and then reseat as pressure is relieved.

However, the FWLB/LOP/SF analysis does not result in the PSVs reopening later in the event and the PSVs do not pass water for which they are not specifically qualified.

In the current FWLB/LOP/SF analysis, the assumption is that the RCP bleed-off rate is a minimum of three gallons per minute throughout the event. However, as noted above, it is more conservative to assume that operators will isolate normal RCP seal bleed-off or there is a failure of the CBO header relief valve to open to the RDT during the FWLB/LOP/SF event, and that the RCP seal bleed-off rate will be zero gallons per minute. If the zero RCP seal bleed-off rate is assumed in the FWLB/LOP/SF analysis, 4

Enclosure Evaluation of the Proposed Change the absence of RCP seal bleed-off may result in a higher peak pressurizer level for the FWLB/LOP/SF event. The additional RCS mass, combined with the normal RCS heat-up, may cause the PSVs to open a second time, later in the event, when the pressurizer water level would be above the PSV nozzles. This would cause the PSVs to pass water which does not meet the PSV design basis acceptance criteria.

To ensure that the PSVs continue to meet their acceptance criterion and not pass water, the FWLB/LOP/SF event analysis was revised (Reference 6.4). In the revision, the event timeline remains the same for the associated SSCs with the exception of the operator action. Specifically, the revised analysis credits the operators with opening an ADV to control RCS heat-up and pressurizer level rise at 20 minutes (1200 seconds).

This specific method of cooling is not mandated as there are others available, but its equivalent effects are credited in the analysis for control of RCS heat-up rate and pressurizer level increase.

Operator Action The FWLB/LOP/SF event analysis is revised such that operator action is now credited after 20 minutes (1200 seconds) of the transient instead of the previous 30-minute (1800 seconds) time frame. The current and the revised analyses are supported by the current PVNGS EOPs which contain explicit operator instructions to ensure that the plant is placed in a stable, safe condition following FWLB events.

In the revised analysis, the credited operator action at 20 minutes after event initiation is to open an ADV on the unaffected steam generator. This action controls the RCS heat-up and prevents the PSVs from opening a second time to preclude a potential challenge to the RCS inventory control and RCS heat removal safety functions. To illustrate the affects of operator action at 20 minutes on the nuclear steam supply system (NSSS),

the revised FWLB/LOP/SF event analysis continues for the remaining 10 minutes of the 30-minute transient, and demonstrates acceptable control of RCS heat-up and pressurizer level increase, with no challenge to the integrity of the NSSS.

Consistent with the revised analysis, the current PVNGS EOPs direct the operators to maintain the pressurizer level below the level of the PSVs throughout the event and provide the operators with variable methods to perform that action (e.g., opening the ADVs, securing charging flow, etc.). The operators are directed to take these actions based on pressurizer level, not time. As a result, crediting operator action at 20 minutes does not change the directions to the operator provided by the EOPs. It only assumes that the directed actions will be completed as required during the event. As discussed below, 20 minutes is considered by the industry as a reasonable length of time for the operators to determine what has happened, assess the current conditions, and initiate required actions to mitigate the conditions. Since the PVNGS EOPs are not time sensitive, they do not require a revision as a result of the revision to the FWLB/LOP/SF analysis.

The assumption of operator action within 20 minutes after the first few alarms are triggered is based on ANSI/ANS Standard 58.8, "Time Response Design Criteria for Nuclear Safety Related Operator Actions" (Reference 6.5). The ANSI/ANS Standard 5

Enclosure Evaluation of the Proposed Change states that for Plant Conditions IV and V, the minimum time that operator action can be credited to occur is 20 minutes from the time of the alarm until initiation of safety functions. ANSI N18.2, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants" (Reference 6.6), and UFSAR Chapters 15.0.1.3 and 15.2.8.4.1, describe the FWLB event as a Limiting Fault/ Condition IV event. Based on the ANSI/ANS Standard 58.8 guidance, an operator action at 20 minutes is acceptable.

For this change in credited operator action, the guidance provided in Example 4 of Section 4.3.2 of NEI 96-07, Revision 1, "Nuclear Energy Institute Guidelines for 10 CFR 50.59 Implementation" (Reference 6.8), was followed. Example 4 illustrations provide cases where changes would be acceptable as they would not result in more than a minimal increase in the likelihood of a malfunction of a SSC important to safety for a change in operator action time.

The change in credited operator action time to open an ADV on the unaffected SG to 20 minutes (from 30 minutes) also meets the guidance provided in NRC Information Notice (IN) 97-78, "Nuclear Energy Institute Guidelines for 10 CFR 50.59 Implementation" (Reference 6.9). The various areas of guidance and how APS meets them are discussed below:

, Procedures and Training Verify that the action (including requiredcompletion time) is reflected in plant procedures and operatortrainingprograms.

The action being taken, (i.e., opening of the ADV on an unaffected SG to preclude a direct challenge to the RCS inventory control and RCS heat removal safety functions) is currently listed in the EOP procedures for standard post trip actions, excess steam demand, loss of all feedwater, and loss of offsite power / loss of forced circulation events (References 6.10, 6.11, 6.12, and 6.13, respectively). The operators are trained on all EOP procedures.

The EOPs contain explicit instructions to ensure that the plant is placed in a stable, safe condition following a FWLB event. These procedures direct the operators to maintain pressurizer level below the level of the PSVs throughout the event and provide them with variable methods to perform that action (i.e., opening the ADVs, securing charging flow, etc.). The operators are directed to take these actions based on pressurizer level and not time. Crediting operator action at 20 minutes does not change the directions for the operator or methodologies provided by the EOPs. It only assumes that the directed actions will be completed as required during the event. In accordance with ANSI/ANS Standard 58.8, 20 minutes is a reasonable length of time for the operators to determine what has happened, assess the current conditions, and initiate required actions to mitigate the conditions. Since the EOPs are not time sensitive, they do not require a revision as a result of the revision to the FWLB/LOP/SF analysis.

APS commits that upon NRC approval of this proposed amendment, the operator action completion time will be added to the Palo Verde Time Critical Action Program (Reference 6.14).

6

Enclosure Evaluation of the Proposed Change Verification of Performance Demonstrate that the action can be completed, in the time required, consideringthe aggregateaffects, such as workload or environmental conditions, expected to exist when the action is required.

The assumption of operator action 20 minutes after event initiation is based on ANSI/ANS Standard 58.8. The ANSI/ANS Standard states that for Plant Conditions IV and V, the minimum time that operator action may be considered to occur is 20 minutes from the time of the event indication. This guidance is based on a reasonable response time for that event. The EOPs controlling recovery for these events are not affected by the change in the credited time for the operator action in the safety analysis. Those EOPs have been previously verified by simulation to support the required actions to control RCS heat-up and pressurizer level when mitigating these events under aggregate affects. There are no known normal or emergency environmental or equipment conditions that would adversely affect the completion of the actions in the EOPs within the credited 20 minutes.

• Error Recovery Considerthe ability to recover from credible errors in performance of manual actions and the expected time requiredto make such a recovery.

There are three potential operator errors with respect to the operator action taken being incorrect or omitted:

1. Operator Opens the Wrong ADV:

The sequence of events for a feedwater line break demonstrates that by approximately two minutes into the event both the MSIS and the auxiliary feedwater actuation signal (AFAS) have been initiated and have performed their respective design functions. In addition, the AFAS logic has actuated to prevent addition of feedwater to the affected SG, based on the SG differential pressures (DP), and is only providing auxiliary feedwater (AFW) to the unaffected SG. This is commonly referred to as the AFW DP lockout or AFW SG high DP lockout feature. After approximately five minutes from the event initiation, the affected SG is empty and at atmospheric pressure, while the MSSVs on the unaffected SG continue to cycle providing sufficient evidence for the operators to recognize the event and determine which SG is affected, based on SG pressure, AFW flow and SG level. Based on this, assuming the operators open the ADV on the wrong (affected) SG at 20 minutes after initiation of the event is not considered credible.

2. Operator Manipulates the ADV in Wrong Direction:

The ADVs are normally in the fully closed position and require operator action to initiate opening them. Manipulation of an ADV switch in the wrong direction (i.e.,

closing the ADV) has no effect.

7

Enclosure Evaluation of the Proposed Change

3. Operator Takes No Action:

The revised analysis credits operator action at 20 minutes to ensure control of the RCS heat-up and pressurizer level rise. All of the related EOPs require the operators to control RCS heat-up and pressurizer level as a basic priority in response to FWLB events. The operators are trained on that requirement and the failure of an operator to provide this control is not considered credible. All the associated plant instrumentation and alarms are provided to assist the operator to quickly identify and manage the FWLB event. Although specific use of the ADVs to control RCS heat-up is not a requirement of the EOPs, control of the RCS heat-up and pressurizer level is required.

None of the potential operator errors are considered credible; therefore, error recovery is not a concern. If an error were to occur for any reason there is ample indication of plant status available to the operators to ensure timely recovery.

• Effects on Plant Systems Considerthe effect of the change on plant systems.

The change in operator action time does not introduce a change to plant systems since the operator action (i.e., opening an ADV) is currently specified in the EOPs.

The UFSAR in Section 10.3.2.2.4 lists component design of the ADVS as:

"Atmospheric dump valves, one per main steam line, are provided to allow cooldown of the steam generators when the main steam isolation valves are closed, or when the main condenser is not available as a heat sink."

The steam generators are the primary. heat sink for the RCS. As a result, use of an ADV to cooldown a SG'results in cooling of the RCS and control of the pressurizer level: The change in operator action time does not change the design functions of the ADVs, SGs, or the RCS. The change merely assumes the design functions take place earlier in the event response. This ensures that the pressurizer level is controlled and the PSVs remain operable throughout the event. This has always been the design function of these systems and this is not changed or affected by this change in the assumed operator action time requirement.

The revised FWLB/LOP/SF analysis evaluates the transient for 30 minutes, including the remaining ten minutes after operator action takes place, to demonstrate the effect the operator action has on the NSSS. That analysis demonstrates that pressurizer level will be maintained below the PSV level and no water will pass through these valves. In addition, the revised analysis demonstrates acceptable affects on the NSSS from the earlier operator action throughout the event.

Conclusion The proposed reduction in the credited operator action time in the revised FWLB/LOP/SF analysis is acceptable based on the guidance of ANSI/ANS-58.8.

8

Enclosure Evaluation of the Proposed Change Based on operator actions at 20 minutes, the revised analysis shows that the pressurizer level will be maintained below the PSV level and no water will pass through these valves throughout the 30 minute FWLB/LOP/SF transient. It also demonstrates that there is no adverse affect on related SSCs.

This change does not involve a change to the type or amount of any effluents that may be released offsite, or increase the individual or cumulative occupational radiation exposure. This change ensures that the licensing bases limits are not affected and remain bounding.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix A Criterion 15--Reactor coolant system design The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that thedesign conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

This change in the accident analysis assumption of operator action within 20 minutes does not affect the reactor coolant system design or operation. The EOPs provide adequate direction to the operators to ensure that the reactor coolant system continues to meet the requirements of Criterion 15.

Criterion34--Residual heat removal A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

The EOPs provide adequate direction to ensure that the pressurizer level is maintained below the PSV level and the pressurizer does not go water-solid during the FWLB/LOP/SF transient. This maintains the residual heat removal capability of the RCS.ý This change in the accident analysis assumption of operator action within 20 minutes does not affect the maintenance of those parameters.

4.2 Precedent None 4.3 No Significant Hazards Consideration Determination The proposed license amendment would change the feedwater line break with loss of offsite power and single failure (FWLB/LOP/SF) analysis described in the Palo Verde Nuclear Generating Station (PVNGS) UFSAR, by crediting operator action at 20 minutes into the event to control reactor coolant system (RCS) heat-up and pressurizer level.

9

Enclosure Evaluation of the Proposed Change The current analysis assumes no operator action until 30 minutes into the event at which time operators would initiate a cooldown of the reactor coolant system (RCS)'per plant procedures. The proposed change is the result of a lack of a design basis analyzed value to use for the reactor coolant system (RCS) bleed-off rate assumed in the FWLB/LOP/SF event analysis, which has resulted in Arizona Public Service Company (APS) conservatively setting the bleed-off rate at zero gallons per minute.

The use of that assumption in the FWLB/LOP/SF analysis could challenge the continued operability of the pressurizer safety valves (PSVs) due to overfilling of the pressurizer. The proposed change credits the operator action at 20 minutes to control RCS heat-up and pressurizer level and then at 30 minutes, credits initiation of a cooldown of the RCS in accordance with plant procedures.

APS has evaluated continued operation and determined that the condition addressed by this proposed amendment does not pose an operability concern., This is based on an existing flowpath from reactor coolant pump bleed-off line to the reactor drain tank, not credited in the FWLB/LOP/SF analysis, that will reduce the potential rate of level increase in the pressurizer. In addition, the emergency operating procedures (EOPs) contain explicit directions to help the operator ensure that the plant is placed in a stable, safe condition following a feedwater line break (FWLB) event. Those directions include actions to be taken to control RCS heat-up and maintain the pressurizer level below a level that could pass water through the PSVs and potentially challenge RCS inventory control. The EOP operator actions are parameter dependent and not time dependent, and crediting them at 20 minutes is reasonable, acceptable, and consistent with the guidance in ANSI/ANS-58.8, "Time Response Design Criteria for Nuclear Safety Related Operator Actions," dated 1984. The proposed change does not affect the EOP directions for continued control of RCS heat-up and pressurizer level increase during the FWLB/LOP/SF event, and ensure that the required safety functions continue to be met.

APS has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change in the credited operator action time to, 20 minutes from 30 minutes does not change the probability of a FWLB/LOP/SF event as the operator actions are credited after the start of the event.

This change in operator action time does not adversely affect accident initiators or precursors, the ability of structures, systems, and components (SSCs) to perform their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits, or radiological release assumptions used in evaluating the consequences of an accident previously evaluated.

10

Enclosure Evaluation of the Proposed Change Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change in the credited operator action time to 20 minutes from 30 minutes does not involve any design or physical changes to the facility or any SSC of that facility. The proposed change does not create any new failure modes or adversely affect the interaction between any structure, system or component. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change in the credited operator action time to 20 minutes from 30 minutes does not alter the manner in which safety limits or limiting safety system settings are determined. No changes to instrument/system actuation setpoints are involved. The safety analysis acceptance criteria are not impacted by this change and the proposed change will not permit plant operation in a configuration outside the design basis. The assumed 20 minutes for operator action is consistent with Industry and NRC guidance. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, APS concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact 11

Enclosure Evaluation of the Proposed Change statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 "CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," CENPD-107, April 1974 6.2 "Final Safety Evaluation for Topical Report WCAP-1 5996-P, Technical Description Manual for the CENTS Code" (TAC No. MB6982), Letter, Herbert N.

Berkow (NRC) to Mr. Gordon Bischoff (Westinghouse), dated December 1, 2003 6.3 NRC letter, "Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments, RE: Various Administrative Controls," dated October 15, 2001 (ADAMS Accession No.ML012880473) 6.4 TA-13-C00-2002-002, Revision 5, "Pressurizer Safety Valve Operability and Long Term Decay Heat Removal Analysis Based on the Feedwater Line Break Event."

6.5 ANSI/ANS-58.8, "Time Response Design Criteria for Nuclear Safety Related Operator Actions," 1984 6.6 ANSI N18.2, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," 1973 6.7 ANSI/ANS-51.1, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," 1983 6.8 NEI 96-07, Revision 1, "Nuclear Energy Institute Guidelines for 10 CFR 50.59 Implementation," November 2000 6.9 NRC Information Notice 97-78, "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, including Response Times", dated October 23,1997 6.10 Procedure 40EP-9EO01, "Standard Post Trip Actions" 6.11 Procedure 40EP-9EO05, "Excess Steam Demand" 6.12 Procedure 40EP-9EO06, "Loss of All Feedwater" 6.13 Procedure 40EP-9EO07, "Loss of Offsite Power / Loss of Forced Circulation" 6.14 Procedure 40DP-9ZZ04, "Time Critical Action (TCA) Program" 12

ATTACHMENT UFSAR Chapter 15 Page Markups Pages:

15.2-66 15.2-68 15.2-70 15.2-72 15.2-76 15.2-78 15.2-79 Figure 15.2.8-42 Figure 15.2.8-43 Figure 15.2.8-44 Figure 15.2.8-45 Figure 15.2.8-46 Figure 15.2.8-47 Figure 15.2.8-48 Figure 15.2.8-49 Figure 15.2.8-50 Figure 15.2.8-51

PVNGS UPbATED FSAR DECREASE IN HEAT REMOVAL BY THE SECQNDARY SYSTEM relieve RCS pressure. The maximum liquid volume attained in the pressurizer during the FWLB event remains below the inlets RpqcT to the PSV nozzles when they are open (the PSV nozzles are at and 1738 ft pressurizer 13 an elevation equivalent to 99.4% level volume). Thus, the pressurizer does not go solid at any time and RCS pressure control is maintained. The transient is terminated at 1800 seconds, when operators initiate a controlled cooldown, such as by using ADVs, to shutdown cooling entry conditions. The operator can take action to isolate the affected steam generator and refill the unaffected steam generator by manual control of AFW any time after the reactor trip occurs. However, the FWLB with LOP and Single Failure analysis does not credit any operator action for the first 4 O minutes of the transient.

Analytical setpoints and response times associated with the Reactor Protection System (RPS) trip functions and Engineered Safety Features Actuation System (ESFAS) functions were chosen to be consistent with, or conservative with respect to, limiting numerical values that appear in the PVNGS Technical Specifications and UFSAR Chapter 7.

The NSSS is protected during this transient by the primary safety valves (PSVs) and the following trips:

• Steam Generator Low Level

" Steam Generator Low Pressure

" High Pressurizer Pressure

• Low Departure from Nucleate Boiling Ratio (DNBR)

• High Containment Pressure

• Variable Overpower Trip.

June 2005 15.2-68 Revision 13

PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY

1~C-, IAr ýl r nr%!S r-T Cr THE SECONDARY SYSTEM TABLE 15.2.8-5 SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK WITH LOSS OF OFFSITE POWER AND SINGLE FAILURE EVENT Time sec.) Event RTF NRTP 3876 N90

.0 0.0 FWLB occurs; A complete loss of norma feedwater 0.0 0.0 to both steam generators occurs 26.1 29.4 AFAS generated in unaffected stear/generator 26.8 29.9 Pr surizer pressure reaches HPP setpoint1 26.8 29.9 HPP signal generated; SIAS/CI /MSIS signal gener ed 26.8 30.0 PSVs op Affected \team generator ries out; AFAS generated 27.0 30.2 in affected steam generiytor 27.3 30.4 Reactor trip reakers /pen 27.3 30.4 Turbine trip o curs/

27.5 30.7 Maximum RCS pre- u occurs 27.9 31.0 Scram CEAs begin Alling 30.3 33.4 LOP occurs 32.4 35.6 MSIVs close 34.6 35.6 MSSV bank 1ipens (firt time) 34.9 38.3 PSVs close/

36.2 41.5 AFW lock t occurs 37.8 37.2 MSSV b k 2 opens 37.8 41.9 Stea/generator pressure reachs a maximum 50.3 58.1 MSS bank 2 closes 70.3 73.5 0 charging pump re-starts 73.0 76.2 FW initiated to unaffected steam knerator 78.4 85.9 MSSV bank 1 closes 487.0 --- / PSVs open 488.6 - PSVs close 1792.0 1 0.0 Maximum liquid volume of pressurizer occui 1800.0 800.0 Operator initiates plant cooldown The HPPT is coincident with the LSGLT.

June 2005 15.2-70 Revision 13

PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM CIAS/SIAS/MSIS are initiated on high containment pressure at the time of reactor trip. This methodology does not change the timing of the reactor trip, or the methodology for matching the HPPT with dryout of the affected steam generator. Parametric analysis determined that the time of reactor trip is the most adverse time to initiate the CIAS/SIAS/MSIS. This methodology assumes that a CIAS/SIAS/MSIS occurs on high containment pressure, simultaneously with the high pressurizer pressure trip. Early MSIS (before SG low pressure occurs.) is conservative with respect to pressurizer level criteria since it eliminates the unaffected SG cooldown through the break early in the transient. There is no effect on peak RCS pressure due to early MSIS, since the peak pressure occurs before the closure of the main steam isolation valves (MSIVs).

A SIAS causes the charging pumps to load sequence to the diesel generator after LOP, depending on demand from the PLCS. This is a conservative assumption for pressurizer fill criteria since it adds inventory to the RCS.

The pressurizer Level Control-System is in the -

automatic mode with the plant operated on program Tavg at the start of the transient. This methodology provides justification for using the nominal cold leg temperature as the initial cold leg temperature assumption for the event.

edJJ TINS..'T D ýr June 2005 15.2-72 Revision 13

PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY Rr[qa r-n-Ce. Tct6L THE SECONDARY SYSTEM

' T. T it -Ft E Table 15.2.8-6 ASSUMED INITIAL CONDITIONS FOR FEEDWATER LINE BREAK WITH LOS OF OFFSITE POWER AND SINGLE FAILURE EVENT Value /

RTP RTP Parameter 3876 3990 MWt / MWt It I. - $

Initial coreRower (% of RTP) 102 / 102 Initial averag RCS temperature, Tavg (at 100% power a maximum pressurizer 583.0 585.6 level) (OF)

Initial pressurizer essure (psia) 2100 2100 Initial RCS flow (% de ign) Z 95 95 Initial pressurizer wat level (%) 59% 59%

Initial steam generator w er level 25% (70% WR) 25% (69% WR)

(%NR) wl Moderator Temperature Coefficent y 0.0 0.0 (x10-4 Ap/OF)

Fuel Temperature Coefficient Least Least negative negative Kinetics Maximum

• Maximump CEA worth of trip-WRSO(p) 8.0 8.0 Scram delay time (sec) 0.5 0.5 CEA Holding Coil Del time (sec) 0.6 0.6 Fuel rod gap conduc ance (Btu/h-ft 2 -OF) 00 500 Plugged steam genrator tubes 1000 500 0/1258 PSV Opening Set mt (psia) 2475-\ 2475 PSV setpoint Vlerance -1% -1%

PSV rated fl w per valve (lbm/hr) 460,000 460,000 PSV Blowdo 14.2% 14.2%

MSSV set mt tolerance +3%

MSSV B/wdown 5%

Singl,OtFailure One AFW Pump One A W Pump LOP,' YesOYe2%

Ffedwater line break area, ft2 0.24 0.

/

June 2007 15.2-76 Revision 14 LDCR 07-F016

PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM performance for this limiting fault event are identical to those described in UFSAR Section 15.2.8.4.3.B.

C. Results The response of key RCS parameters as a function of time is presented in Figures 15.2.8-42 through 15.2.8-51 for this limiting fault event.

The limiting peak pressure FWLB events are discussed in Sections 15.2.8.2 and 15.2.8.3.

For FWLB with LOP and a single failure, auxiliary feedwater actuation in the affected steam generator is delayed until affected steam generator dry-out. A main steam isolation signal on high containment pressure isolates the unaffected steam generator from the break early in the transient. Following the isolation, AFW delivery increases the level in the unaffected steam provides sufficient i nventory for-heat removal to occur through the MSSVs _suc~h that RCS pressure control is .

maintained-by the PSVs, and the R CS converges to a quasi steady state prior to 1800 seconds (Figure 15.2.8-44 and

46) . Operator action may be taken at 1800 seconds. This demonstrates the adequacy of RCS decay heat removal with the AFW system during the FWLB which satisfies SRP 10.4.9 and 15.2.8.

Throughout the transient (Figure 15.2.8-46) the pressurizer water level remains below the PSV inlet nozzle location of 99 .4% level or 1738 ft' volume, any time the PSVs are open, and only steam is discharged as required by June 2005 15.2-78 Revision 13

PVNGS UPDATED FSAR DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.8.4.5 Radiological Conseauences and Containment Performance Fuel damage is not predicted for this limiting fault event.

The dose consequences for this event are no more limiting than the dose consequence assessment presented in section 15.2.8.2.5.

15.2.8.4.6 Conclusions The auxiliary feedwater capacity is adequate to provide removal of the core decay heat until operator action is taken59.4 minutes after event initiation.

The maximum pressurizer water level remains below the PSV inlet nozzles and only steam is discharged, thereby satisfying NUREG-0737 Requirements presented in UFSAR Chapter 5B and 18.II.D.

June 2005 15.2-79 Revision 13

j;RIE LACWITH NEWF~ IGURE  :

I. . . ... . .. . . .. . . . . . . . J 3876 MW 3990 MW

.96 72 ri 48 0

0J 24 0

0 600 900 .1200 1B00 TIM!E, seconds PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE CORE POWER vs. TIME FIGURE 15.2.8-42 JUNE 2005 REVISION 13

JREPLACE WITH-NEW FGUR.

Tht D- 612 588 564 so0 0 600 900, 1800 flME, sonds

RELCWTH NWFGJE.

1800 Un 1200 600 0

0 600 900 1 1800 TIME, seconds PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE RCS PRESSURE vs. TIME FIGURE 15.2.8-44 JUNE 2005 REVISION 13

[REPLACE WITH NEW FIGURE 2400 CA1800 Ix1200 600 a

0 600 900 1200 1800 TIME, seconds PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE PRESSURIZER PRESSURE vs. TIME FIGURE 15.2.8-45 JUNE 2005 REVISION 13

IREpLAC__ WITH NEFW FI!GU RE 1 191EPLACENWITH{ NEW FGR

! . . .. . . . .. . ." . .. . .. . .. . .. .. . .... . ... ... .. . . ... ..... I 1320

• ) Unaffected SG 990 0

0 600 900 1200 1500 1800 TIME, seconds \

PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE SG PRESSURE vs. TIME FIGURE 15.2.8-47 JUNE 2005 REVISION 13

I IýfL9WITH NEW FIGURE`.ýl I

--- - --- 3876 MW 3990 Mw 108000 0

z 720 36000 0

0 TI0 900 1200 1800 TIME, seconds PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE SG LIQUID INVENTORIES vs. TIME FIGURE 15.2.8-48 JUNE 2005 REVISION 13

............................ I IREPLACEW WITH NE:W FIGURE

-.- . . . . .-.. 3876'M W

-l m s 060

1. iL 270 0

z I-S 90 0 300 600 90 22 .1500 1800 TIME, seconds PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE AFFECTED SG AFW FLOW vs. TIME FIGURE 15.2.8-49 JUNE 200.5 REVISION 13

jfREPLACE WIT-. ýN.E.WýýFIGUPR I 270 re 0ý 180

ýz 90 0

0 600 900 1200 J800 TIME, seconds PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE UNAFFECTED SG AFW FLOW vs..TIME FIGURE 15.2.&-50 JUNE 2005 REVISION 13

IREPLACE WITH. NEW!ýFIGUREý 540

.30 180 -

0 0 300 600 900 1200 1500 1800 TIME, seconds PALO VERDE NUCLEAR GENERATING STATION UPDATED FSAR FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE PSV FLOW vs. TIME FIGURE 15.2.8-51 JUNE 2005 REVISION 13

INSERT A Depending on the feedwater line break size, a containment pressure high trip and a containment high-high pressure Engineered Safety Features Actuation System (ESFAS) signal may also occur for an inside containment FWLB. In order to account for the containment pressure trip condition and coincident containment isolation actuation signal (CIAS), safety injection actuation signal (SIAS) and main steam isolation signal (MSIS), it is assumed that a containment pressure trip condition and CIAS/SIAS/MSIS occur at the same time as the high pressurizer pressure trip. In addition, the containment spray actuation signal (CSAS) is assumed to activate coincident to the reactor trip and resulting ESFAS signals.

INSERT B The maximum liquid volume attained in the pressurizer during the FWLB event remains below the volume which results in water entrainment into the PSV nozzles when they are open (the PSV nozzles are at an elevation equivalent to 99.4% level).

Insert C TABLE 15.2.8-5 SEQUENCE OF EVENTS FOR FEEDWATER LINE BREAK WITH LOSS OF OFFSITE POWER AND SINGLE FAILURE EVENT Time EVENT Value (sec.)

0 FWLB and complete LOFW to both SGs, break size (ft'). 0.199 31.07 Pressurizer pressure reaches trip setpoint (psia.) 2450 HPPT signal generated. SIAS/CIAS/MSIS/CSAS signals 31.07 generated 31.15 PSVs open (psia). 2450 31.22 Dryout of affected SG (lbm of liquid inventory), <5000 AFAS generated in affected SG.

31.57 Reactor trip breakers open.

31.57 Turbine trip occurs 31.87 Maximum RCS pressure (psia) . 2572 32.18 Scram CEAs begin falling.

34.57 LOP occurs 36.69 Main Steam Isolation Valves close 37.26 MSSVs bank 1 open (psia)

• 1303 38.77 MSSVs bank 2 open (psia) 1344 39.32 PSVs close (psia) 2102 43.19 Maximum SG Pressure (psia) 1367 43.96 AFW Lockout (psid). 270 59.64 MSSVs bank 2 close 1276 74.57 One charging pump restarts (gpm) 44 77.24 AFW initiated to SG #2 (one pump, gpm). 650 91.31 MSSVs bank 1 close (psia) 1237 1200 Operator Action - Intact SG ADV opened (%) 10 1800 Maximum liquid volume of pressurizer (ft'). 1683 1800 Operators initiate plant cool down Notes: 1 - The HPPT is coincident with the LSGLT 2 - Only the first time the Bank 1 MSSV open and close is listed. The Bank 1 MSSVs continue to cycle as required to remove decay heat.

INSERT D

• CSAS initiated on high-high containment pressure at the time of reactor trip. This methodology does not change thetiming of the reactor trip, or the methodology for matching the HPPT with dryout of the affected steam generator. It is conservative to assume that a Containment Spray Actuation Signal (CSAS) coincident with the CIAS signal. Since the containment peak pressure and temperature analysis did not explicitly analyze any feedwater line break events, [Chapter 6, Section 6.2.1.1.1.1] the exact amount of time that elapses from when the containment pressure reaches the high containment pressure trip setpoint to when it reaches the high-high containment pressure trip setpoint is unknown. Therefore it is conservative to assume that it reaches the high-high containment pressure trip setpoint at the same instant it reaches the high containment pressure trip setpoint. A review of Chapter 6, Table 6.2.4-2 determined that the activation of the CSAS isolates the CBO return to the VCT, Instrument air, and nuclear cooling water penetrations into / out of containment and opens the spray header isolation valves. For this analysis, the key effect is the isolating of the CBO return to the VCT. This is a conservative assumption for pressurizer fill criteria since it adds inventory to the RCS.

• The PVNGS Emergency Operating Procedures (EOPs) contain explicit instructions to help the operator manage to ensure that the plant is placed in a stable, safe condition following an Excessive Steam Demand event. Therefore, the analysis assumes operator action to open an ADV (on the intact steam generator) to preclude a direct challenge to the RCS Inventory Control and RCS Pressure Control Safety Functions twenty (20) minutes after the event initiation.

INSERT E Table 15.2.8-6 ASSUMED INITIAL CONDITIONS FOR FEEDWATER LINE BREAK WITH LOSS OF OFFSITE POWER AND SINGLE FAILURE EVENT Parameter Value Initial Core Power (% of rated) 102 Initial average RCS temperature, Tavg (at 100% 585.6 power and maximum pressurizer level) ('F)

Initial Pressurizer Pressure (psia) 2100 Initial RCS Flow (% of design) 95 Initial Pressurizer Water Level (ft) 23.9 Initial SG Water Level (ft) 35.8 (25% NR) 4 -0.2 Moderator Temperature Coefficient (xl0- Ap/OF)

Fuel Temperature Coefficient least negative Kinetics Maximum CEA Worth at Trip-WRSO (%Ap) 8.0 2-°F) 500 Fuel Gap Gas Conductance (Btu/hr-ft SCRAM delay time (sec) 0.5 CEA Holding Coil Delay Time (sec) 0.6 Plugged SG Tubes 0%/10%

PSV Tolerance -1%

PSV Blowdown 14.2%

MSSV Tolerance +3%

MSSV Blowdown 5%

Single Failure One AFW pump LOP yes FWLB Area (ft') 0.199 RCP seal control bleed-off flow rate (gpm/RCP) 0.0 Operator Action Time (minutes) 20.0 INSERT F The AFW flow provides sufficient inventory for heat removal to occur through the MSSVs such that RCS pressure control is maintained by the PSVs, and the RCS converges to a quasi steady state prior to 1200 seconds (Figure 15.2.8-44 and 46). Operator action is taken at 1200 seconds. This demonstrates the adequacy of RCS decay heat removal with the AFW system during the FWLB which satisfies SRP 10.4.9 and 15.2.8.

Throughout the transient (Figure 15.2.8-46) the pressurizer water level remains low enough so that only steam is discharged from the PSVs any time the PSVs are open, as required by UFSAR Chapter 5B and 18.II.D for meeting the NUREG-0737 Requirements.

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 120 ..........- 1 .

100 80 0

60 0

0 40 20

~1 I I I

I I 1.11.11111 0

0 300 600 900 1200 1500 1800 TIME, seconds CORE POWER vs. TIME FIGURE 15.2.8-42

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 660 640 620 rj 600 Q

580 560 540 0 360 720 1080 1440 1800 TIME, seconds UNAFFECTED LOOP RCS TEMPERATURES vs. TIME FIGURE 15.2.8-43

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 3000 2500 2000 1500 1000 500 0 L-0 300 600 900 1200 1500 1800 TIME, seconds RCS PRESSURE vs. TIME FIGURE 15.2.8-44

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 300 0 . . . . . . . . . I . . . . . . . . . . . . . . . . . . . .. . . . . . . . .. . .. . . ..

2500 ow 2000 1500 1000 500

0. . . . ... I, ,..............

0 300 600 900 1200 1500 1800 TIME, seconds PRESSURIZER PRESSURE vs. TIME FIGURE 15.2.8-45

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 180 0 .. .. . . .. . .. . . . .. . . . .. . . .. . .]. . .. . . .. . .. . . .. . .. I. . . . .

1500 1200 900 600 300

. .Ill..lll lli l l. .

1... 1I III. I.

0 ......... ........ i ...II 0 300 600 900 1200 1500 1800 TIME, seconds PRESSURIZER WATER VOLUME vs. TIME FIGURE 15.2.8-46

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 1650 1375 1100 0u 825 550 275 0

0 300 600 900 1200 1500 1800 TIME, seconds SG PRESSURE vs. TIME FIGURE 15.2.8-47

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 180000 I-150000 Ci~2 120000 S

90000 z

60000

- Unaffected SG 30000 Affected SG 0

0 300 600 900 1200 1500 1800 TIME, seconds SG LIQUID INVENTORY vs. TIME FIGURE 15.2.8-48

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 450 II 11111,1 un II IIIuIII~IIuIIuI,, liii liii U

U, 375 0

300 0

225 150 75 0

0 300 600 900 1200 1500 1800 TIME, seconds AFFECTED SG AFW FLOW vs. TIME FIGURE 15.2.8-49

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 450 375 300 0 - I I I I II I III 11 11 1 LIII 1 1 1 II II 225 150 z 75 I I I I [~I I~ 11 11 1 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 0

0 300 600 900 1200 1500 1800 TIME, seconds UNAFFECTED SG AFW FLOW vs. TIME FIGURE 15.2.8-50

FWLB WITH LOP AND SINGLE FAILURE LONG TERM COOLING CASE 900 750 600

,0 450 300 150 0

0 300 600 900 1200 1500 1800 TIME, seconds PSV FLOW vs. TIME FIGURE 15.2.8-51