ML051520445

Significant Adverse Condition Investigation Report

ML051520445
Person / Time
Site:	Harris
Issue date:	05/27/2005
From:	- No Known Affiliation
To:	Office of Nuclear Reactor Regulation
References
111308, FOIA/PA-2004-0277
Download: ML051520445 (53)

Text

ATTACHMENT 16

. A>; -Sheet 1. of 2 Significant Adverse Condition Investigation Report Form CAP-NGGC-0205-1-16 Action Request/Nuclear Condition Report Number: 111308 Event Time: 13:34 Facility: RNP Event Date: 11/19/03 Unit: 2 Investigators: Bruce Gerwe, Richard Hightower, Brad Dolan, John Little, John Valentino, Grant Chappell, Frank Modlin, Scott Jackson, Vic Smith I. Event Description During the development and review of Revision 0 of Engineering Chang two postulated fire time critical transient conditions were identified. These conditions, i not mitigated could result in an unrecoverable plant operating condition.

If a postulated fire were to occur causing specific circuit damage, operator actions to mitigate the transient would have to be taken in less than 10 minutes from the onset of circuit damage. Based on current RNP analysis criteria, operator mitigating actions taken outside the control room required in less than 10 minutes are not considered acceptable due to the limited amount of time that the Control Room Staff would have to detect equipment malfunction, determine its effect and then take mitigating actions. The time transient conditions identified include:

r a) A postulated fire event that causes the spurious operation (closing) r

_ _ _ _ _ _ = _ Ok__ ________

9-I .5 This scenario would cause loss of nute and if o eratin at the time, b) A postulated fire event that causes theo ration of boto ClF'Tt .jconcurrently. This scenario potentially would cause loss of an unrecoverable amount of RCS inventory in less than 10 minutes.

Both these time transient conditions could occur given a fire of sufficient magnitude in one of the following two fire zones:

frznsicui h is comprised of 8 additional 1w ~ san ipendix RLH.G.3 safe shutdown fire area, requiring alternate or dedicated shutdown from outside-the Fire Area.

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- 2. Problem Description Problem Statement Two postulated fire induced circuit failure cases have been identified that require the development of new thermal hydraulic cases to determine the potential impact to the Appendix R Safe Shutdown methodology. The identified cases had not previously been analyzed for impact to the Appendix R Safe Shutdown methodology. Mitigating actions (automatic system / component actuations, control room operator actions or manual actions taken outside the control room) must be accomplished before exceeding allowable operating parameters, or before unrecoverable conditions exist. It is likely that operator actions taken to mitigate these events may not be completed rapidly enough to assure safe shutdown can be accomplished as currently evaluated.

For purposes of this discussion, an unrecoverable condition is defined as a condition which results in fuel clad damage, rupture of any primary coolant boundary or rupture of the containment boundary.

Problem Discussion One of the Appendix R Performance Goals requires reactor coolant inventory to be maintained within the indicating range of the pressurizer level instrumentation, and control of reactor cooling system pressure.

)as identified two potential fire induced circuit failure cases, which if not mitigated could compromise this requirement. The postulated conditions are based on the reasonable assumption that two concurrent spurious operations of safe shutdown equipment occur during the event. Appendix R analysis requires the licensee to assume the fire event damages cables in the fire area causing spurious signals to be implanted from one cable to another or from one conductor to another within the same cable. This in turn, results in components being energized and moving to a position opposite of their desired safe shutdown position. Appendix R also requires for m.G.3 fire areas, that a loss of off-site power must be assumed at the onset of the event. However, the loss of off-site power cannot be used to the benefit of the scenario. The postulated cases are described below:

Case No. 1 -

For a postulated fire in Fire Zone Dedicated Shutdown is credited using the n now hich is su lied ower from the Dedicated Shutdown Bus.

urng this event. Should thb running ie to support normal operation, the potential exists to damage to the pump and or associated piping, such that the Reactor Coolant make-up function is not maintained.

Based on the Appendix R Safe Shutdown Apalysi iRM Ml.. i.

ould spuriously o te rsulting in both valves being closed at the same time for a fire in Fire Zone Since Primary Water and Boron Injection Systems are not part of the Appen ix Sae Shutdown Analysis, they are assumed to fail as a result of the postulated fire. The Wre also designed without

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- high discharge or low suction pressure pump protection. During normal operation, any two of the three pumps may be running.

In this postulated case, Off-Site Power is conservatively assumed to be available, supplying power to the Emergency and Dedicated Shutdown Busses. Thee jis I' ctrical power from the Dedicated Shutdown Bus and the are supplied electrical power fro ectively. With off-site power available under thi scenario, a _Alsovith off-site power available and thR ineaffected by the initial consequences of the fire, they could be available foi1 Case No. 2 -

For the second stulated case,_

iaU U puriously open. This case identifies a potential uncontrollable loss of Reactor Coolant Inventory. This is because both of the c4

__arear open and cannot t be closed during the postulated fire event.

_Trre motor operated valves which are supplied electrical power fri supplied electrical power fronimIp= As such, a Loss of Offsite Power must be assumed concurrent with the postulated fire.

Because the postulated fire may damage Emergency Diesel Generator circuits, including the EDG output breakers, the Emergency Diesel Generators cannot be credited for this fire scenario. As such, the normally open VCannot be closed due to the loss of electrical power t ll-Th re closed by de-energizing 125VDC power that supplies their circuits. However, this manual action taken outside the control room does not take place for approximately 10 minutes following entry into the Safe Shutdown pr.Qedures. Based on lessons learned from TMI, following the opening of the

_ I _ Therefore, the operator actions taken to mitigate this event cannot be accomplished in the required time frame.

Consequences No actual fire events or loss of safe shutdown capability have occurred. This NCR deals only with postulated fire events. These postulated fire events could cause the Appendix R performance goals and objectives not to be met. In the case of spurious closing orp-

_g_ potentially lost leading toaote 'a] loss of natural cool down capability. In the case of spurious opening of 'n uncontrollable loss of RCS inventory could occur leading to core uncovery.

Immediate notification was made to the NRC on November 19, 2003. LER 2003-03 is scheduled to be submitted by January 20, 2004.

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Initial Extent of Condition Initially, this condition appeared to be limited to a ostulated fire in either r

• with the spurious closing of both a sa, fthe time. similar A

condition exists in these fire zones for spurious opening of bot..

the same time.

Refer to Section 7 for further evaluation of extent of conditiop. No extension of this condition was found to currently exist beyond Fire Zone Interim Mitigating Strategv Upon discovery of this condition,**was imme iatel vised to take preemptive control room operator action for a fire in Fire Zones! These actions are:

a) toveri reclosed, and b) to vef h snt an operating pump @

Closure of Block Valve . Insures that an open RCS vent path does not initially exist upon concurrent spurious operation (opening) of both Pressurizer PORVs and with a loss of offsite power.

Verification that th s not an operating pump mitigates damage to the pump, maintains the pump water solid and enables the pump to remain available to be used for RCS inventory makeup as needed.

3. Investigation Summary Summary The methodology applied in this investigation is consistent with the requirements of CAP-NGGC-0205. However, due to the nature of the condition, existing techniques are supplemented with a design review of the condition. This design review first determined the Appendix R functional and operating design criteria and requirements, followed by a determination of whether the present design meets these requirements. It was determined that a design deficiency exists between the present design and the design bases requirements.

A Barrier Analysis and Cause and Effect Analysis were conducted to determine cause of failure (See Attachments 2 and 3).

This condition was discovered during the development of he purpose of the Engineering Change is to perform a feasibility analysis of credited manual actions taken outside the control room. The activity was being performed by experienced engineering staff on site, in conjunction with experienced engineering staff of Framatome-ANP.

Framatome-ANP was contracted by Progress Energy to perform the feasibility study and to.

assemble the engineering change. The original Appendix R analysis was performed by outside Vendors knowledgeable in Appendix R and assisted by CP&L personnel knowledgeable in plant systems.

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As a part of the engineering change, certain postulated fire induced hydraulic cases were developed to confirm the engineering requirements. The two cases identified above are examples of some of the cases postulated. This engineering basis is needed in order to prioritize operator actions, ensuring Safe Shutdown Performance goals and objectives are met.

During the latter stages of the development of these cases, it became evident that the two postulated cases above would be problematic. Specifically, that the operator actions may not be achievable in the required time frame to prevent unrecoverable conditions.

Due to the historical nature of this issue, no environmental conditions and potential error precursors could be established.

Appendix R Chronological Time Line 1975 - Browns Ferry Fire 1981 - Appendix R Rule Promulgated 1981 - Generic Letter 81-12, Fire Protection Rule, Issued. First letter providing guidance on circuit analysis 1983 -1985 - RNP SER Submittals / NRC Approval of the Appendix R SER 1986 - Generic Letter 86-10, Implementation of Fire Protection Requirements, Issued.

Provides additional guidance on Appendix R Fire Induced Circuit Failures. This letter was issued after NRC approval of the Appendix R Program at RNP.

2003 - Discovery of Unanalyzed Condition at RNP.

Detailed Review Methodology and Results The design engineering review includes three distinct elements. They are as follows:

1) Review the Appendix R Design and Licensing Basis requirements,

2) Review the current Design and Licensing basis to insure appropriate design considerations are factored into the Safe Shutdown Analysis, and

3) Results I Conclusions of the Design Review.

(1) Review of Design and Licensing Basis to determine requirements:

The Appendix R Safe Shutdown Performance Goals and requirements are described ir"-

he requirements embodied in this document are directly derived from 10CFR50 Appendix R, Section M.L. The Appendix R Safe Shutdown performance goal that is not met as a result of the two postulated fires is as follows:

Performance Goal and Definition Reactor Coolant Makeup - Maintain the reactor coolant inventory within the indicating range of the pressurizer level instrumentation, and control reactor cooling system pressure.

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- (2) Design Review:

For a postulated fire in Fire Zones the Licensing and Design basis for achieving Safe Shutdown conditions are embodied in IOCFR5O, Appendix R, Section LI.G.3. This section requires Alternate or Dedicated Shutdown capability due to the inherent damage to both trains of safety related equipment in these fire zones. A Dedicated Shutdown System is essentially a minimum capability safe shutdown train independent of normal shutdown trains. This is comyeferred to as tb ~ ^

s the operating procedure that provides the necessary operating instructions for achieving Hot Shufdown conditions, for this postulated fire event. For urposes of maintaining reactor coolant inventory credit is taken for starting thee and isolating the' ubsequent to the initiation of the postulated fire.

In this postulated fire events Loss of reactor coolant inventory is to be limited to only reactor coolant pump seal leakage.

This is accomplished by removing power from th thereby causing their closure. The are deactivated (currently within 10 minutes of entry intor o the closed position and only the mechanically-operated relief valves to the ilbe available for primary system pressure relief. ThdtlI

_iw CJ I1 U1*s also blocked by the deenergization of the isolation valve, thereby causing it to close. Additional are blocked closed by de-energizing the system isolation valves.

Following reactor trip, the secondary system will be used to remove decay heat, causing shrinkage of the RCS inventory during cool down. As mentioned above,_

will cause a futher reduction in the primary system inventory. Additonal borated water from thiw-*s added to the system by means ol d Reactor coolant makeup by use of _I In addition, somne of th e is utilized to These functional requirements must be maintained independent of the fire induced circuit failure(s) resulting from the postulated fire. IntJhey__ -

9ir_ also identified as spurious operations concerns.

A review of the safe shutdown supporting mechanical calculations determined that a hydraulic analysis was performed for the potential loss o ,i 'ue to spurious actuations of th

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This calculation determined the impact to the ssuming spurious actuation of the appropriate combination of valves. The results and conclusions of this calculation indicate that the operator actions taken to mitigate the event are performed in an acceptable time frame. The time critical nutes) manual operator action taken outside W-terminates the event, prior to undesirable conditions existing.

For both scenarios, the original analysis assumed that appropriate operator actions could be taken prior to the postulated fire induced circuit failures that would result in operation beyond the allowable operating limit(s). Therefore, no hydraulic analyses were performed or required for some cases where two concurrent spurious operations could be postulated. Review of the original safe shutdown analysis and supporting calculations support this fact. Noprior hydrauilic analysis couldb Ifind to establish the technicals bass contrary to other cases as identified above, where two concurrent spurious operations were considered and hydraulic analyses were performed. Shifting regulatory guidance added to the confusions surrounding the history behind the licensing basis. This resulted in unclear guidance that was not evenly applied.

This apparent assumption in the design was unsubstantiated. It assumed that no fire induced circuit failure(s) would render safe shutdown equipment inoperable, as long as manual i, (1

action could be taken early in the event. The need to perform hydraulic analyses to back up I this assumption was not evenly applied to all cases. This is classified as an "Apparent Assumption", because it is not documented, either in the Safe Shutdown Analysis or supporting licensing correspondence.

Mechanical Engineering therm ulic analyses described i a ave determined the potential exists to lose the f assuming the pump is running during normal operation. Based on this analysis, loss of suction to the pump would occur in less than one minute following closure oN",. As later discussed in the Safety Significance Section, catastrophic failure of the pump is not expected to occur for at least 15 minutes following loss of suction. The pump would be able to function when later called upon, but at a reduced efficiency due to air entrainment. Since the _ _ _ u r n uring this postulated fire, the potential exists that the plant would be in an unrecoverable condition.

As later discussed, the postulated failures result in minimal damage to the This would not preclude thu Im performing its intended design function during the event. Securing the _t the onset of fire in the areas provides a preventative measure to mitigate pump damage, maintain the pump water solid and enable th_

-as needed.

Additionally, the ydraulic case for the excess and normal letdown valves have also been determined in, C Based on the results and conclusions of this analysis, operator actions taken to terminate spurious actuation of the excess and normal letdown valves are accomplished in the required time frame.

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A calculation has been prepared to address the potential fo o be open. This calculation concluded that the reactor core would remain covered within the time it would take for operator actions to remove power to the using them to go closed. This calculation assumed a two phase flow through the The calculation used a simplified model of only the reactor vessel and pressurizer.

It ignored the large volume of fluid in The calculation started with approximately _ v h iie th W 2

'Additionally, a simulator run mimicking the event was performed and showed the event was even less of a concern (See Attachment 5).

The simulator run revealed that only steam flow would result throug less than half the amount of mass loss assumed in the calculation. It also showed the Q01 Afiluring the event swung from a starting point of _

7in NO TheaoSmatru roth resuAL-%lte lnaBt h calculation and simulator run both accounted for reactor head voiding.

All spurious actuation cases that could result in a loss of RCS inventory have now been analyzed inn 1L¶-. -L--M. iI1MjS i J- L1 (3) Results of the Design Review:

Guidance for performance of the original fire induced circuit failures in the Safe Shutdown N; Analysis is unclear. This has a direct effect on the number and type of fire induced circuit failures assumed during a fire event. This in turn, influences the mechanical hydraulic cases which may be required to support Appendix R Performance goals and objectives. For purposes of the hydraulic cases considered, time equals zero at the onset of the postulated fire induced circuit failures. This design deficiency in how spurious operations are applied has existed since the original analysis. Since no prior revalidation effort of the entire original analysis has been undertaken until recently, this deficiency was never discovered.

The current Safe Shutdown Analysis references Generic Letters 81-12, Fire Protection Rule, and 86-10, Implementation of Fire Protectionr Requirements, in the Reference Section, which specified different criteria. The Safe Shutdown Analysis requires that Fire Induced Circuit Failures consider "any and all one at a time". It is not known whether this was intended to be sequential with time to recover from one before proceeding to the next, or if the events were to be postulated concurrently. Contrary to this guidance, examples exists which clearly consider two hot shorts concurrently at the system level. The following is a partial listing: (Note: Each of these cases has been evaluated and is acceptable.)

/ '-'IN

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However, some cases appear to only conside one fire induced fail Ire. An. xample would be the de-energizinurin f, urg a postulated L *re, to preclude the possibility of a The root cause of this event is the unclear guidance provided in the Appendix R Safe Shutdown Analysis on postulated fire induced circuit failures. The corrective action to prevent reoccurrence of the event is to establish clear guidance on the performance of circuit analysis for Safe Shutdown purposes and ensure our program is in alignment with this guidance.

Progress Energy recently completed a position paper on Fire Induced Circuit Failures for Appendix R purposes tobe considered at all Progress Energy Nuclear Operating Facilities.

This paper clearly requires two postulated fire induced circuit failures to be considered at the system level in the Appendix R Revalidation Project at each site.7;fhis position paper clearly defines the number and type of circuit failures to consider. RNP will need to complete reanalysis of the Appendix R Program utilizing the criteria specified in the Progress Energy position paper.

4. Inappropriate Acts / Equipment Failures The inappropriate act is an incomplete design analysis. The inappropriate act involved the original Appendix R analysis team for RNP.

5. Causal Factor Associated with each InappropriateAct / Equipment Malfunction

1. Causal Factor: Unclear design criteria/guidance at the time of the original design analysis.

This is historical as the original analysis was performed in the 1980s.

Cause Code: Ki Type: Root Cause

2. Causal Factor: Lack of technical justification to support criteria/guidance. This is historical as the original analysis was performed in the 1980s.

Cause Code: KI Type: Contributing

6. Previous Operating Experience (Internal and External)

Internal OE CR3 reported in October of 1997 that a design error resulted in inability to provide reactor coolant system inventory makeup during an Appendix R event. A reanalysis of Appendix R Fire Study determined that the power supplies for the high pressure injection valves and the normal inventory makeup valves were not protected from the postulated Appendix R fire in

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the control room or cable spreading room. This resulted from a deficient fire study previously performed. The event was determined to be not significant.

RNP Response: This event occurred before CR3 joined the Progress Energy nuclear fleet.

This event was not released as OE by the utility, and therefore, was not previously evaluated by RNP. As part of the RNP long term plan for continued upgrade and evaluation of its Appendix R analysis, projects were begun in 2003 to address manual action feasibility and those portions of the analysis that would uncover similar problems.

In December of 2002, HNP identified that postulated fires in three fire areas could cause spurious closure of certain valves. Spurious closure of valves in the flow path for the protecte uld result in loss of the protected~U~

if it was in service at the time of the postulated fire. Similarsimultaneous multiple spurious closures of valves in the flow aths of water to ould result in loss o iredited in the SSA and subseq uent Q4 In January of 2003, it was identified that simultaneous multiple s urious opnn of -certain valves could result in transferring of 7_TT cause of these conditions is inadequate original Safe Shutdown Analysis of certain conductor-to-conductor interactions.

RNP Response: As part of the RNP long term plan for continued upgrade and evaluation of its Appendix R analysis, projects were begun in 2003 to address manual action feasibility and those portions of the analysis that would uncover similar problems.

OE A search was conducted of the Nuclear Network for Operating Experience on the subject of this NCR. Only one item could be found that had been posted as an OE item.

In February of 2002, during revalidation of the Appendix R compliance strategy for Indian Point Unit 2 it was discovered that the analysis lacked sufficient detail and/or support documentation to justify the adequacy of separation of the original and current design configuration of th control cables. Panels for control and local/remote breaker control for althree pump controllers are located in a common hallway immediately outside tled ubicles. This configuration leaves all control functions (i.e. both breaker and speed control) for Wulnerable to damage by a single fire event in the area. This issue was determined to be not safety significant.

RNP Response: RNP has had a long term plan for continued upgrade and evaluation of its Appendix R analysis. In 2001 evaluation of manual actions for I.G.3 areas was completed.

Work on manual actions for M.G.2 areas was to begin in 2002, but was postponed until 2003 due to the power up-rate project. Work was begun in 2003 to address the remaining manual actions and those portions of the analysis that would uncover similar problems.

LERs A total of four events concerning r 9were found.in the LER database.

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1. In September of 1998 at St. Lucie Unit 2, it was determined that the cables providing control signals to the vere not adequately isolated from adjoining cables and could cause the o spuriously open in the event of a control room fire. A fire isolation switch rovided for this circuit; however, it did not addeuately isolate all portions of the control circuit. The station attributed thecable issue to misapplication or misinterpretation of NRC requirements for circuit failure analyses.

This event was determined to be not significant.

2. In October of 1999 at Salem Unit 1 a concern was identified with the cable routing of the during a review of the post fire safe shutdown 6na yis. Thecab-le for each

_MA-Mamisisow-' is routed in the same cable tray inside the containment. In the event of a postulated fire inside containment =Wo uud lose power and can not b closed. The fire cIuld also cause a hot short to occur that would cause the associate o spurously open. These two occurrences together would result in the The cause of the event was the failure to properly evaluate the interface function of th during the development of the Appendix R safe shutdown analysis. This event was considered to be not significant.

3. In May of 2000, Prairie Island Unit 2 determined that th yfj in containment do not meet the Appendix R 20-feet separation criteria. The scpnario of concern is a fire in containment causing a sustained external short circuit in th circuitry that would result in the concurrent with the same fire causing a ground or open circuit in the circuitry that would prevents - With these two valves opened, ,a_ S.-uld occur that could not be isolated. The cause of the event was an oversight with considering th wwhen the safe shutdown analysis was updated to implement atheguidance of NRC Generic Letter 86-10. This event was considered to be not significant.

4. In April of 2002, Millstone Unit 3 determined that the emergency operating procedure (EOP) did not provide adequate assurance that thFMO would be disabled within a timeframe that would prevent inadvertent actuation-from a fire-induced short circuit an NOR

_~l a I ee assumed that 15 minutes. wpl povid Uhd adequate time to isolt th and establish positive control fnt te event of a fire ini However, reanalysis of operator actions and response times could not assure isolation of th Assumptions to support the fire safe shutdown analysis were not properly validated. This event was considered to be not significant.

RNP Response to each: These LERs were not released by the utilities as Operating Experience; therefore, they were not evaluated by RNP. These events are being addressed within this NCR.

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7. Extent of Condition The loss o C_ *e-,oration (spurious closin fn has been addresse i eview of cable routing f orl ound no other lant fire zones where a similar problem exists. On the discharge side ofonly one electrical motor operated valve is in the flow ath. Motor operated valve ,isthe,,There is no impact to r component operation should this valve spuriously close. . Therefore, _Jdoes not impact the ability to achieve safe shutdown conditions.

The spurious actuation hydraulic cases associated with a loss of, ave been completed. These cases were evaluated in vjat These analyses IFZ9%FNIFW 'Lr - , :-;wMrdrff:~~S ..... I ,,1 address the folloionn~i I -7 Review of cable routings fork u und that they are also routed in other fire zones; however, int ese cases other mitigating factors, such as the availability of SI, preclude them from being a concern.

Several other fluid systems function to support Appendix R safe shutdown. They are Auxiliary Feedwater, Component Cooling Water, Service Water and the Residual Heat Removal Systems. These systems were evaluated in the original analysis for fire induced circuit failures under the same apparent assumption that manual actions could be taken early in the event prior to circuit failures resulting in unrecoverable conditions. However, these systems / components need to be reevaluated to determine if there are postulated fire induced circuit failures that could result in unrecoverable conditions. The extent of condition cannot be fully determined until such an analysis is complete. This is an extensive analysis, which will require significant engineering effort to complete.

This engineering evaluation is currently part of the overall Appendix R Safe Shutdown Analysis Revalidation Project, which is being completed by Sargent and Lundy for RNP.

The task is currently scheduled for completion by July, 2004 at RNP. This task will be relied upon to complete the extent of condition for the nuclear condition report.

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Appendix R Safe Shutdown Analysis Revalidation Project - Treatment of Spurious Operation of Eqcuipment The Appendix R Safe Shutdown Analysis Revalidation Project is currently an ongoing project. Project Instructions for Task 5 of this effort, specifically addresses the methodology and treatment of spurious operation of equipment. These instructions state that two spurious concurrent mal-operations of equipment must be considered in the circuit analysis for cables and equipment being credited for shutdown of the plant. Project Instructions for Task 6 will identify all hydraulic analyses that must be considered. Plant modifications are anticipated to be required following completion of the revalidation project.

8. Safety Significance Summarv This issue deals with the possibility that fires in certain locations in Fire Zone could caus kroscoue

__1 his would result in loss of suction to A concern was raised that tis could potentiall result in rapid damage to and loss of the twou There ar but approximately two-thirds of the time, the be in service and t us it could potentially be subject to damage. Th is important in the safe shutdown analysis because it is powered from the dedicated shutdown bus.

It is important to note that there has been no actual failure associated with the issues identified in NCR 111308. In addition, no actual fire initiating event is being reported by the NCR that challenges the operability of the charging pumps. The only question examined in this assessment is whether there was a significantly higher likelihood of core damage at some time in the past due, to the external events I fire issues identified in this NCR than had previously been understood. The following provides this historical review of the event prior to the establishment of the. compensatory action to take preemptive control room action to verify the This compensatory action maintains the pump water solid. - --

Vendor information (See Attachment 4) and an event at St. Lucie (described below) support the ability of positive displacement charging pumps to run for a limited period of time without suction and not cause sufficient damage to the operating pumps that would prevent them from being restarted. The head of water developed froml

_jT is head will help to push air from the pumps. Local control in th

'_is available to the operators to start and control the pumps as needed. Since only two of the three pumps are operating at the start of the event, the non-running pump remains full of water and not entrained with air. Once the supply is restored to the pump(s), they will be capable of supplying water into the system. Should the non-ninning pump be available to be started, normal flow into the system is expected since

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the pump remained full of water. If one of the previously operating pumps is started, then a forward flow of water is still expected. Dedicated shutdown (fire response)

-d plign e Therefore, on a qualitative basis this potential concern would not result in any significant increase in risk in comparison with prior estimates.

This issue identified concerned the possibility that fires in certain locations in Fire Zones

,could cause concurrent spurious opening of botLN uting in an, I before operators could take mitigating actions.

It is again important to note that there has been no actual failure associated with the issues identified in NCR 111308. In addition, no actual fire initiating event is being reported by this NCR that challenges the ability of the PORVs to operate or be isolated as required. The only question examined in this assessment is whether there was a significantly higher likelihood of core damage at some time in the past due to the external events / fire issues identified in this NCR than had previously been understood.

It has been determined through calculation and a simulator ru thtmi ating actions currentl Proceduralized in the dedicated shutdown procedure to isolate J f1_~re adequate to prevent core damagewhether one or two PORVs spuriously open: Therefore, this new potential concern would not result in any significant increase in estimates of risk in comparison with prior estimates.

The issue o ~purious operation was previously addressed by the NRC and Progress EnergI. In lI ter NLS-85-0732, dated 11/21/85,.he NRC approved the methodology to clos arly in the fire event. The NRC stated they found this method of "ensuring prevention of fire induced spurious operations of these applicable high/low pressure interface valves acceptable." As such, the dedicated safe shutdown procedures provide a means by which this is accomplished.

Additionally, Operators are routinely trained on this procedure.

Maintaining thiin the closed position has been considered, but has been determined to be not preferable at this time. Maintaining the in the closed position during normal operation would prevent the

_jWfr from relieving pressure durin an.L This could result in unnecessary actuation of th I uring a transient condition. By maintaining the atic control and unblocked condition, it is less likely that th will be tuated. This is considered to be preferable, because the availability of the both the nd the

.j~pprovides additional mitigation capability for an and reduces the likelihood of malfunction of th valves.

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Initial mitigating action to close rom the control room is not a concern if during the event should they spuriously operate to the open position. First, the valves are closed to prevent a loss of power event from preventing their closure. Once closed, a spurious operation would constitute a third and fourth spurious operation. This is beyond our current and redefined (Appendix R revalidation effort design basis.

Additionally, it has been shown that even with bo _after manual actions are taken to remove power fo h y It are in series, so closure of either one removes the leakage path.

Impact of Loss of Suction to Charging Pumps A scenario has been postulated which would result in the for approximately 15 minutes while ex eriencing a loss of suction. This wou d be due to he con cer with this scenario is a catastrophic ran ure of the operating pump(s). In addition, after a period of approximately 36 minutes the lawwould be called upon to provid WM-- e \1CamN When s suction is lost, air entrainment and associated cavitation in the pump will be experienced rom both dissolved gases in the liquid as well as air residing in the

__@ uction stabilizer. RNP actual experience has shown that such air entrainment iv-oZld cause cavitation within the pump resulting in pitting and increased wear on the l alve train components. This observed wear was over an extended period of time; on the order of two to three months. The wear was also seen to accelerate based on the operating speed of the pump. At that time RNP was operating with only one i Iwhich _ thereby increased the speed of the pump, the cavitation, and subsequently the wear. Operating two pumps at lower speeds decreased this phenomenon significantly and eliminated the i te wear completely. RNP currently practices dual pump operation for the_ reduce maintenance costs and wear.

However, onl _l eured to meet system needs.

In this scenario, with the continued operation of the pumps, cavitation would peak as more and more air is introduced into the pump. Eventually, the hydraulic forces acting on the internal valve components would begin to decrease due to a reduction in hydraulic forces caused by the replacement of water with air. There would be an increase in associated pump operating noise accompanying this timeline, as well as associated increased wear on the power end components due to unbalanced forces within the fluid cylinders of the pump.

This would reach the same threshold as the internal valve components, followed by a reduction as more and more air enters the pump. The other obvious correlation is the loss of pumping efficiency as more air enters the pump. It should be noted that the pumps do not rely on the pumped fluid for lubrication of power end or motive force components other than for some cooling of the packing which again would result in some accelerated wear. This wear is not expected to be significant as the primary lubrication and cooling for the packing is provided by the ubrication tank supplying primary water.

Operating experience from IN 83-77 discusses some events at nuclear power plants involving pumps failing to function due to gas entrainment. One of those events involved

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positive displacement "and the discussion of the event supports the conclusion that positive displacement charging pumps can withstand a period of operation without a suction supply:

St. Lucie On October 23, 1982, with St. Lucie Unit I in hot standby during recoveryfrom a reactor trip, the three operatingpositive displacement chargingpumps stopped injecting coolant to the reactorcoolant system because the volume control tank (VCT) was pumped dry.

The reactorhad tripped on a low steam generatorwater level signal after a loss of feedwaterflow to the steam generator.T7e VCTiwas empty although its two liquid level sensors indicated an acceptable liquid inventory and hence an apparentlyacceptable inflow/outflow balancefrom the VCT. The hydrogen cover-gas blanket of the VCT entered the suction of each pumnp. Thefalse liquid level indication was caused by an empty reference leg that was sharedby both liquid level sensors. The pumips wvere restored to operation by repeated venting afterfilling the VCT to a high level.

Based on RNP experience in internal valve train and packing wear, characteristics and principles of o peration of the pumps, and previous operating experience, a catastrophic failure of thel s not expected to occur while operating with a loss of suction for approximraely'15 minutes. Ten minutes is considered a reasonable time eriod for the operators to recognize the situation and stop the operatingThe vendor, contacted to provide their insights into this scenario as well.

They came to the same conclusion; the pumps would not be expected to experience a catastrophic failure during this time period. Their reply is provided in Attachment 4.

Impact of Fire Protection Fire Zones 'e protected by a full area fire detection system and automatic Halon suppression system. The detection system for each fire zone is provided with two trains of detection. Actuation of both trains of detection will cause the Halon system to automatically discharge. The Halon system is a total flooding suppression system consisting of a main and reserve bank of cylinders. This redundancy in suppression capability alsominimizes system out of service time. Detection system design information identifies that for each fire zone, the number of detectors actually installed exceeds the required minimum number of detectors. Upon actuation of a fire alarm by one of the detection trains, the Control Room will initiate an investigation and appropriate fire brigade response.

Procedural guidance already in place requires that upon the loss of either the fire detection/actuation system or the fire suppression system for the affected fire zones, a continuous fire watch will be put i lcj__

The penn nt fire loading in Fire Zone is considered "high", while the fire loading-in Fire Zone s considered "moderate". Existing plant procedures contribute to the fire safety of the plant by controlling the use and storage of combustibles, maintaining housekeeping standards nd controlling sources of ignition. In addition, non-qualified IEEE-383 cables in Fire Zone re coated with a fire retardant that will slow down the propagation of fire between circuits.

- 16

Fire ZonesIM are adjacent to each other. Access into Fire Zones from Fire Zone g iFour unannounced fire drills have been conducted in these fire zones in the last two years. Fire drill response times for each fire zone show an operator on the scene between 1 and 7 minutes. This response time starts when the initial alarm is received in the control room. In each of these drills, the fire brigade was on the scene between 10 and 13 minutes from the sounding of the plant fire alarm.

Cable Testing!- Times to Failure and Spurious Operation EPRI document 1003326, Characterization of Fire-Induced Circuit Faults - Results of Cable Fire Testing, discusses spurious actuation of devices in electrical circuits due to fire induced damage to electrical cables. This document includes recent fire testing of circuits. Section 12.2.5 gives results of the time to cable failure and Section 12.2.6 provides information on spurious actuation. For the types of cables originally installed at RNP (thermoplastic), the test results give an average time to cable failure in 15 minutes. The average time to spurious actuation is 25 minutes, with an average spurious duration of less than 3 minutes. These values support the fact that on average, sufficient time exists for the execution of manual actions prior to initiation of spurious operations and that when spurious operations occur, they are short in duration.

Time Lines Generc Tihe Line Assumptions

1. Automatic fire suppression system does not operate or extinguish the fire.

2. No Operator or fire brigade actions are taken in the fire zone to control or extinguish the fire.

3. A maximum of two spurious operations of components occur at the onset of the event.

4. Entry into the dedicated shutdown procedures (DSP) begins with the first spurious signal.

5. First and second spurious signals occur concurrently, followed by subsequent failures.

Subsequent failures are taken one at a time.

6. Spurious operation of equipment has no specific end time.

7. Loss of off-site power is required to be taken when it can do the most harm and cannot be used to benefit the fire scenario.

8. Credit cannot be taken for cables or components that have not been analyzed for Appendix R.

-Sm-Time from Start Description of Action of Scenario (Minutes) r.

(Mintes Fire starts in FZ6 [\A _

X

• First and second spurious operations clb s 4_

- 17

MitigatingFactorsNot Taken Credit ForIn Charging Pumip LCV Scenario

1. First and second train fire detection signals being received in Control Room.

2. Automatic Halon system actuation following detection of fire by the second train of detectors.

3. Built-in redundancy of detection and suppression systems: two detection trains, two Halon actuation circuits, and main and reserve banks of Halon cylinders.

4. Operator investigation initiated with first fire alarm received in Control Room. Four ri1, recent unannounced fire drills reported investigator on the scene between 1 and 7 minutes.

5. Fire Brigade on the scene between 10 and 13 minutes as recorded in the four recent unannounced fire drills.

6. No large single fire source exists in either Fire Zone

7. Non-IEEE-383 qualified cables in Fire Zone re coated with a fire retardant that will slow down the propagation of fire between circuits.

8. Average duration for spurious signal operation as determined by fire testing is three minutes.

9. Loss of off-site power is not taken in this event. If taken vould stop running, further minimizing any damage to the operating pumps.

10. The CVC System design. The head of water developed from th bove the elevation of the pumps, all provide for a positive suction su ply to the pumps. This head will help to push air from the pumps. Local control in thd_"<

mI3is available to the operators to start and control the pumps as needed. Since only two of the three pumps are operating at the start of the event, the non-running pump remains full of water and not entrained with air. Piping to each pump can be vented locally at its suction stabilizer upstream of the pump. If off-site power has not been lost and the non-running pump is not affected by the initial consequences of the fire, it could be started.

11. If SI is not affected by the initial consequences of the fire, and if is needed, it could be

_I started.

- 18

Mitigating Factors Not Taken CreditForIn k y _ rio

1. Same as the items 1 through 8 above for th ccnario.

2. Scenario is analogous to Station Blackout Event. C-Conclusion Based on the above, it is highly unlikely that these events would have lead to unrecoverable conditions. Loss of bot toould have resulted in minimal damage to the operating nor to manual actions taken to restore the flow path. Therefore, there would be no significant increase in risk from this event.

In the case of , it has been shown by calculation and a simulator run that the manual actions to remove power and close the valves can be achieved prior to core uncovery with no resulting core damaged. The simulator run shows the conservative nature of the calculation with respect to water remaining above the fuel. Therefore, there would be no significant increase in'-risk from this event.

Cable testing shows that on average, sufficient time exists for the execution of manual

• actions prior to initiation of spurious operations and that when spurious operations occur, they are short in duration.

The design of the fire detection and Halon suppression systems, lack of a significant fire source in either fire zone, fire retardant cable coatings and fire brigade performance make it

- 19

highly unlikely that a fire would not be extinguished or at least controlled to its place of origin. Thereby, limiting fire spread and cable damage to only those initially involved.

With the existence of the detection and suppression system limiting fire spread, it is also unlikely that fire dama would prevent automatic SI initiation immediately following

.l SI operation will stabilize the event.

The compensatory chancyes made i i o add preemptive control room operator actions for a fire in Fire Zones provide the plant with the ability to cope with these postulated fire events both now and as an interim step. These preemptive actions provide sufficient protection until a cohesive plan is in place to deal with these and any future issues identified by the ongoing Appendix R Safe Shutdown Analysis revalidation.

1 In conclusion, based on current plant physical configuration (and not accounting for the interim compensatory measures), both scenarios are outside the required Appendix R design requirements. However, it has been shown that based on actual plant responses to these events (both procedurally and in non-Appendix R credited equipment/responses) that there should have been no significant increase in risk from either of these events.

9. Corrective Action Plan Causal Planned/Completed Actions Assignment Assignee / Initial Due Factor (Annotate Committed Assignments as Type Concurred Date Committed) By CORR Complete 11/19/03

- Operations Night Order 03-024 was CORR Complete released 11/19/03 direptjjithe review of the changes made tc1 Fire 12/04/03 Emergency, by each Operating Shift.

To reduce exposure to the potential effects CORR Complete of a fire from transient combustible materials, the administrative available 11/19/03 limits for the affected fire zones were reduced to 50 percent of the normal allowed loadings.

- 20

Personnel responsible for Appendix R CORR Complete.

Program and circuit analysis (F. Modlin, R. All were Hightower and B. Gerwe) are cognizant of NCR NCR findings and requirement that two Team concurrent spurious mal-operations are to Members.

be considered in Appendix R analyses. 12/23/03 Confirmation that Task 5 of the current CAPR Frank Complete Appendix R Analysis Revalidation process Modlin 12/12/03 includes the assumption that two spurious and concurrent mal-operations are to be considered in the analysis.

Provide written guidance in appropriate CAPR Frank 3/14/04 design documents for performing Appendix Modlin R Analyses that address circuit analysis failures and manual action feasibility.

Guidance to include direction on how to address spurious operation issues and the feasibility to perform manual actions.

Revise NGGC "Fire Protection Technical CAPR Jeff 4/19/04 Position Paper on Fire Induced Circuit Ertman Failure - Circuit Analysis" to reference this NCR and link this NCR with the appropriate steps in position paper requiring that two concurrent spurious mal-operations are to be considered in Appendix R analyses.

Confirm and include the results of this NCR CORR Frank 4/16/04 investigation into an Engineering Document Modlin Task 6 of the current Appendix R Analysis CORR Frank 7/1/04 Revalidation process is to evaluate the Modlin Reactor Coolant System, Auxiliary Feedwater System, Component Cooling Water System, Service Water System, Residual Heat Removal System and their components to determine if there are postulated fire induced circuit failures that could result in unrecoverable conditions.

2 Complete necessa t m ic as CORR Frank 12/1/05 a p e)kModlin (RO-23F)

W 06- ensure spurious operations do not lead to unrecoverable +

events and confirm the extent of condition for this event has been adequately addressed.

- 21

Perform an effectiveness review of the EREV Bruce 4/1/06 completed corrective actions (if effectiveness Gerwe review is waived, provide basis below). *+

• The corrective actions wvill not be completed in the required 120 days due to the complexity and scope of the engineering involved in determining the existence of other unrecoverable hydraulic conditions and proposed resolution(s) to any problems found.

The design activities to correct problems found must be completed in a logical sequence which wvill also impact the ability to complete the corrective actions in the 120 days specified limit.

+ RO-23 Modifications are anticipated to be required.

10. Basis, If Effectiveness Review is waived:

11. PNSC/CSERB Review Required? YES X ] NO D]

. Refer to applicable Implementing procedure

- 22

ATTACHMENT 1 REFERENCE LISTING AR 111308111308The following are the primary references reviewed in the investigation of this Nuclear Condition Report. This listing does not contain all of the documents, drawings, licensing correspondence reviewed in the course of this investigation.

1. Code of Federal Regulations, Title 10, Part 50, Appendix R: "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979.

2. H. B. Robinson Steam Electric Plant, Unit No. 2, Appendix R, Section 111.G Supplemental Submittal; CP&L Letter No. NLS-84-030 dated February 6, 1984.

3. H. B. Robinson Steam Electric Plant, Unit No. 2, Follow-up Clarifications: CP&L Letter No. NLS-84-220, Dated June 6, 1984.

4. H. B. Robinson Steam Electric Plant, Unit No. 2, Appendix R - Alternate Shutdown Capability: Open Item Resolution and Additional Clarification: CP&L Letter No. NLS-85-140, dated June 18, 1985.

5. H. B. Robinson Steam Electric Plant, Unit No. 2, Supplemental Safety Evaluation Report for Appendix R to I0CFR50, Items 11.G.3 and J.L, dated Auguit'8, 1984.

6. H. B. Robinson Steam Electric Plant, Unit No. 2, Supplemental Safety Evaluation Report for Appendix R to 10CFR50, Items 111.G.3 and M.L, dated November 21, 1985

7. H. B. Robinson Steam Electric Plant, Unit No. 2, Appendix R, Alternate Shutdown Capability, CP&L Letter to NRC, Serial: NLS-84-434, November 30, 1984.

8. H. B. Robinson Steam Electric Plant, Unit No. 2, Appendix R Exemption Request, CP&L Letter to NRC, Serial: NLS 85-026. February 13, 1985.

19. GL 81-12, Fire Protection Rule, dated 2/20/1981

20. GL 86-10, Implementation of Fire Protection Requirements, dated 4/24/1986

21. H. B. Robinson Steam Electric Plant, Unit No. 2, Fire Protection Modifications -

Additional Information, CP&L Letter to NRC, Serial: GD-79-871, April 2, 1979.

- 23

ATTACHMENT 2 Barrier Analysis AR 111308111308I

iST1EET11 BARRIER CATEGORY Identify the If a barrier category contributed to the event

• Were any physical barriers not barriers that then assess the specific barrier as follows:

functioning as designed? will be Barrieris No barrierin Barrier was circumvented

• Were there any barriers that did not assessed deficient or place or incorrectly applied perform their functions? I _I failed

-, .>. - f7- '

E.M--- - v fl-, fl'Vfl2h

.-<-. . - -. et fl'

.". -. - ~- a. -a*

Vt".  :

.s -vv:-lTvan

., va,ad** 1-i;v - -,-_ <- . es-;,-

'-- 4_' -n r-z

-s 2 -!.. *:

Design Codes/Standards Unsubstantiated Assumption X caused barrier to be circumvented.

Drawing/Dimensions Other

-- -.'-;;;-;'--.-PHYSICAU BARRIERS=-- -'. :- '.-.- :. ; --.

.Z7--- *-: .

Engineered Safety Features l l l __

Safety and relief devices Conservative design allowances X -. Unsubstantiated Assumption caused barrier to be circumvented. Hydraulic analyses were not prepared to support assumption time line.

Redundant equipment Alarms and annunciators  ;

Fire barriers and seals Other.

PROG RAM CONTROLMONITORING BARRIERS '. 6 .

Training Program l l l _ __

Engineering System Monitoring Human Performance .

Procedure & Document Management Maintenance Rule Lessons Learned Programs Self Evaluation & Assessment Corrective Action Other - Performance of Appendix R Safe X X Unclear criteria/guidance at Shutdown Analysis the time of original analysis.

-~ -  ; .-ADMINISTRATIVEBARRIERS

-- -r Plant Policies & procedures Training and education Equipment Clearances Radiation Work permits Qualification of welders Methods of communication Certification of engineers Regulations Supervisory practices ALARA Other Table is NOT Intended to be ALL inclusive

- 24

T (c4

- 25

J

- 26

Attachment 4 AR 111308111308From: Peck, Larry [1]

Sent: Wednesday, December 17, 2003 3:16 PM To: Little, John Cc: Woods, Paul

Subject:

FW:

John Little Per our phone conversation there is a scenario of complete loss of suction pressure for 15 Minutes. During the loss of suction the pump would G Ax experience cavitation and the accompanying slamming of plungers into pockets of fluid. At some point the pumping chambers would gas bind and the slamming and vibration would cease. We do not expect to see catastrophic failure of the pumps or components due to the incident. There may be additional wear on the load carrying components, i.e., gears, bearings, crank and crossheads. Packing, stuffing box bushings, valves, seats, and springs may also see damage.

Lawrence A. Peck Nuclear Project Engineer

- 27

MPS: ttontfored F4 - Fi1S ryv D S V ot rriable Current Lirmits

1. NJame Value Low High Units Type Description 53.241 0 lE0 pPt none 53.241 0 . 100 pot none 53.241 0 100 pet none 38.552 0 100 pet none 53.2411 0 100 pet none

-24 .0563 0 120 pct none

-4.99607 66 120 pct none

-24.0563 0 120 pet none

-4.99607 66 120 pat none 2235.12 800 2300 psig none 2235.12 0oo 2300 psig none 2235.12 BO0 2300 psig none 75.0588 0 100 pet none 75.058S 0 100 pet none 15 THCELL(241 2257.41 800 2300 psia nor.e CELL PRFSSURE 16 TIHPCELL[21] 2273.07 0OO 2300 psia nore CELL PRESSURZ 17 THPCELL1541 2252.47 800 2300 psia nor.e CELL PRESSURE 18 THWCELL[541 1 .53891 -200 200 lbin/s none CELL MIXTURE FLJ PATE 20 PRTFLlL'Qt41 0 0 100 lb/s nonre LIQUID (VI 21 PRTFLlLItQ51 0 0 100 Ib/s none LIQUID (V) 22 PRTFLlGAS[41 0 0 100 lb/s none GAS (V) 23 PRTFL1GASCSt 0 0 100 lbWS none GAS (V) 24 THALFCR[11 0 0 1 dnlrs none REACTOR CORE VOID FRACTION PASSE 25 TIIALFCRP21 0 0 1 dr.ls none REACTOR CORE VOID FRACTION PASSE 26 THALFCR131 0 0 1 dmlS none REACTOR CORE 'OID FRPNAT10O PASSE 27 THALFCRt4) 0 0 1 daIs none REACTOR CORE VOID FPACTICI PASSE 25 TSHtSTMRCS 3987.2 4000 4001 ibm none PRIMARY RCS STEM4 ASS 29 THMLIQRCS 384750 0 385000 lbm none PRIMARY RCS LIQUID MASS

,4

- 28

mPS: Moritored Parameery u MiP Variable Current Limits i Namie Value Low Moigb:Unitv Type Description 0 0 100 pct none `l

53. 0927 0 100 pct nene 53.0927 0 100 pcE none 0 0 100 pct none

53. 1297 0 100 pet none

-21.779 0 120 pct none

-4.99649 66 120 pet non~e I£4

-21.779 0 120 pr none

-4.99647 66 120 pct none 1700 800 2300 psig none 2231.89 800 2300 psig none 2231.89 D00 2300 psig none 0 0 100 pet r.one 0 0 100 pcz none 15 T-PCELL1241 2252.53 800 2300 psia none CELL PRESSURE 16 T$FPCELL[211 2265.69 600 2300 psia none CELL PRESSURE 17 THPCELL[541 2247.59 800 2300 psia none CELL PRESSURE 18 Th-dCELLE541 150.002 -200 200 lbw./s none CELL MIXTURE FLOW n.ATE 20 PRTFL1LIQ[41 0 0 100 lbis r.one LIQUID (Vl 21 PRTFL1LIQ151 0 0 100 lbis none LIQUID tV) 22 FRTFL1GAS(4) 0 0 100 lb/s r.one GAS (V) 23 PP.TFLGAStSJ 0 0 100 lb/s none GAS (V) 24 THALFCRIlI 0 0 I dmls none f-REACTOR CORE VOID FRACTIOt PASSE 25 TMALFCRE2j 0 0 1 dmls none REACTOR CORE VOID FRACT.ON PASSE 26 THALwCR[31 0 0 1 d~l s none REACTOR CORE VOID FRACTION PASSE 27 27kLC.CRf41 0 0 1 drnls none REACTOP CORE VOID FRACTION PASSE 28 TMMSTMRCS 3938.11 4000 4001 i bm none PRIMARY RCS STEAMt MASS 29 THMLIQRCS 384756 0 385000 1a none PRIARP.Y RCS LIQUID lIASS

_t §I F

- 29

.7 fufjl LI T'- II sr,

( 7 r kA b '.I jiPS: kMani'tored Param-eter Surz1aryI 142 Variable Current Limits 4 flame value Low fitghiUnits Type Description.

2177.68 800 2300 psi.a r.one CELL PRESSURE Boo L

16 TI1RPC~'t2Ij 2187.91 2300 psia none CELL PRESS5mE Boo 800 pe ia 2172 .9 2300 none CELL PRESSURE 12 T1KWCELLtS4I 436.703 -200 200 1bm's nrone CELL IxTURE r'LOC'.Y RATE 20 PRTPFL1.LIQ14J 0 0 100 12,/s none LIQUID (V1 21 PRTFLILXQ[51 0 0 100 1bIs none LIQUID (V1 22 PRTFFLIGAS(41 53 .5694 0 100 ibIs none GAS IV1 23 PRTFL1GASI5J 53.5692 0 100 lI/a none GAS IV) 24 TIIALPCR I 1 0 0 1 dlsi none -REACTOR CORE VOID FACTION FASSE 25 V&[ALFCR(2J 0 0 d Is none REACTOR CORE VOID FRACTIO1 PASSE 1

26 TALFCA131 0 0 dmils none REACTOR CORE VOID FPACTION P.SSE 1

27 TH!ALFCP44j 0 0 d.'sls nsme REACTOR CORE VOID FRACTION PASSE 28 TflMS2'UIACS 4013.02 4000 4001 ibm nor.e PRIIVARY RZS STE.MI 14ASS 29 TIHtLIORCS 3E4608 0 385000 ibm none PRIMARY FCS LIQUID HASS

. v %j (11

- 30

HPS: Monitored Parameter Summary , , ,I 1!P Variable Current Limits f Name valne Low HIgh Units Type Description kT5 THCELL[241 1460.97 800 2300 psia none CELL PRESSURE 16 THPCELL1211 1466.24 800 2300 none CELL PRESSURE psia.

17 THPCELLI541 1456.86 800 2300 psia none CELL PRESSURE 1B '11*CELL1541 -254.237 -200 200 lbm! s nor.e CEL.L MIXTURE FLOW RATE 20 PRTFLlLIQ14] 0 0 100 lbts none LIQUID CV) 21 PRTFL1LIQ15) 0 0 100 none LIOUID IV) 22 PRTFLlGAS[41 33.1897 0 100 none GAS (V) 23 PRTF"lGASCS] 33.1897 0 100 lb/s none GAS 1V) 24 THALFCR[l] 0 0 1 dmls none REACTOR CORE VOID FRACTION PASSE 25 TKULFCR(2I 0 0 1 dmls none -REACTOR CORE VOID FFACTION PASSE 26 THALFCRC3I 0 0 1 dmls none REACTOR CORE VOID FPACTIOt PASSE 22 T1i0LPCRE4 0.000123295 0 1 dmls none REACTOR CORE VOID FRACTION PASSE 28 THSTM1RCS 3014.3 4000 4001 Ibm none PRI4ARY RCS STEAM MASS 29 TIMLIQRCS 379641 0 385000 lbm none PRIMARY RCS LIQUID MASS r/-AM-,. S-' -Q 41g.

- 31

• . Y.Ps. Monitored Parani.~ter SummnrY MP variable Current Limi-ts I1reValu~e Lw igOnsType DescriptiOn.

16

'- THtPCZtLL241 THE'CELL12l1 1062.72 1067.83 800 800 2300 2300 psia psin none none CELL PRESSURE CELL PRESSURE PAS 17 TUPCEL~tA41 1057.02 80o 2300 psla none CELL PRESSURE 18 TfrAwELL 0:4i -34 .5073 -200 200 Ibm3 none CELL MIXTURE FLOW PATE 20 PFRTFLILIQ[41 a 0 100 lb!s none LtQUID W 21 PRTFLlLIQESI 0 0 100 Ibis none LIQUID lV1 22 PR.TEL2GASE41 23.2207 0 100 lbis none GAS MV) 23 PRTFLICAS151 23.2207 0 100 lbis none GAS (V) 24 2T{ALFCRI11 1. 24186e-07 0 1&LI s none REACTOR CORE VOID FRACTIO04 PASSE 25 T1=.FCR[21 0.0524183 0 1dmls none N:ACTCR CORE VOID FRACTIOt PASSE 26 T1{ALFCRC3I 0.0785172 0 1idLmls none REACTOR CORE VOID FPACTION PASSE 27 S14ALFCRC41 0.0648001 0 idnls none REACTOR CCRE VOID FRACTION PASSE 28 Tr4M.STZIRCS 3880.8 4000 4001 Ibm none P?.nm1,RY RCS STEAM WLSS 29 r1{IiLIQRCS 351309 0 385000 Ibm none PRIKARY RCS LIQUID U3SS NW-a~

- 32

()~7 SJAPPR2(Mon Dec 8 16:28:59 2003)

*r.

l47 50 ...........-

40 .. .

10 ....................... i............................ ........................... ................................................................................

20.

0.................

0 100 200 300 400 500 600 Time (Secs) fel o e to L L°°O/i,3 A6 . / 6/

- 33

SJAPPR2(Mon Dec 8 16:28:59 2003) 90 . .. . . . .........

80 .1 .

8 0 . .......... .... ....... ....... .... ................ ........... ........ ...... .....

70 . . . ............ ...

40 ......................... ........................... ............................ .........................

4... ....................... .....

60 ...-

3 0 0 ..............................................................-............................................................. ........... ..........

0 100 200 300 400 500 600 71 Time (Secs)

Or

- 34

a N M - SJAPPR2(Mon Dec 8 16:28:59 2003) 90 ... . . ...

90 . ...................... =..... ................................ ........... ...... ........... ...

70 . .......... t<<.............4.,.,.

70

...... * ................................................... j.............. ...................................................................... ... ...

.60 . ............ .

40 L_

I _ _ _

300 C 0 1 00 200 . 300 400 500 600 7t Time (Secs)

AeTY. 5C . E'/ty

- 35

L4LI

- SJAPPR2(Mon Dec 8 16:28:59 2003) 9..0...... .......

80 . ... . .. .... .

70.

60 _.

40 h................T .......... .............

4 . X.

0 100 200 300 400 500 600 Time (Secs)

_ -rc _I

- 36

~SJAPPR2(Mon Dec 8 16:28:59 2003)

-10

\\, -12 _~.e .e.e.*.e.v

.0.8 e.*.

• .o.oo....._.....v

-14

-16

-18

-20 ............. .. ............... ..... ........... . ........... . ....

-22

-I

-24 0 100 200 300 400 500 600 700 Time (Secs)

IfAZ., s HI O/.z

- 37

SJAPPR2(Mon Dec 8 16:28:59 2003)

-1.5 I -2 I

i

-2.5 i

I

-3 ............. ~~~~I.-......................

-3.5

............. _ _ _ _ I _ _ _ i _ _ _ _

-4 I -4.5 I

-5 0 100 200 300 400 500 600 700 Time (Sees)

A.9

- 38

SJAPPR2(Mon Dec 8 16:28:59 2003)

-10 . .................................................................................................................. .............................. . ......... . ......

ri -12 . ..........................

I

-14 .................................. ............................ ....................

-16 ............................ ............................ ............................ ..........................

-18 . .................. .... .......................................................... ....... ...... ........ ........................................................... .

I

-20 ......................... ............................ ............................ ............................ .. .......................

-22 ......................... ............................ ......................................................... ............................ ............................ ..........................

I,..

-24 .......................... ............................ ............................

0 100 200 300 400 500 600 700 Time (Secs)

S-

- 39

i_ ORi SJAPPR2(Mon Dec 8 16:28:59 2003)

-1.5 .............................. ............................. 17 ........ .......

-2 ............................. ..................................

-2.5 ......... . ............... ............................ ............................ ............................ ............................ .......................... .

-3 .......................... .................. ...... ............................ ............................ ....................................................... .

Ij

-3.5 ............................... ..... ............................................... .... .......... .................... ......................... ...........................

-4 ......................... ............................ .......................... . .......................... .

-4.5 ......................... ..... ................. ............................ ............................ .......................................................... .. . ...................... .

l

-5 ... ............................................. ...... ............................. . ............................

0 100 200 300 400 500 600 700 Time (Secs)

As;\ %I- 13/KlCJ,

- 40

SJAPPR2(Mon Dec 8 16:28:59 2003) 2 .2............

0 .. .. ........ ...... .............. .

2100 .-

2000 ...

1900 .......................... ............................ ............................. ............................ ......................................................... ..........................

1800 .-.

1700 .........

0 100 200 300 400 500 600 7-Time (Secs) f.:h , 1§lo Ifs~ d Lo/g

- 41

SJAPPR2(Mon Dec 8 16:28:59 2003) 2000 . .

1900 .... . . ..

l 1800 .................

1700 _ .... _ ......................

0 100 200 300 400 500 600 700 Time (Secs) dffscALjE Loc At i70 ,

turf

- 42

r _N m t SJAPPR2(Mon Dec 8 16:28:59 2003) 2200 Ij 2100 2000 I. . ........................

1900 i

i 1800 1700 0 100 200 300 400 500 600 700 Time (Secs) 4

_l4 C 1.SCAI6 L.vJ& wr 1J7o tasq 5/. <

Ad. '

- 43

SJAPPR2(Mon Dec 8 16:28:59 2003)

THPCELL[24]

2000 ............. L..

Ad1800 . .

w C,,

U,

.1600.

w 1400 .... .. .

1200 r . ....................... i................ . ............... ............ . . . .............................. ..................... .........................

1000 ., i 0 100 200 300 400 500 600 Time (Secs) i~cn.~

- 44

iiww SJAPPR2(Mon bec 8 16:28:59 2003)

THPCELL(21J 2200 2000 Ca.

w 1800 Er C',

C,,

w cc 1600 0.L

-J

-J .... . .. .. .... . .. .. . . . . ... . . ... . .. .. . . .... .... ..... . .

w 0

1400 1200 1000 0 100 200 300 400 500 600 700 Time (Secs) 7 . .5 Age 1 /R 6

- 45

AMI - SJAPPR2(Mon Dec 8 16:28:59 2003)

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- 47

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- 53