3F0398-16, Forwards Response to 980202 RAI Re Summary Rept on Resolution of USI A-46, Seismic Qualification of Equipment in Operating Npps. Confirmatory self-assessment of USI A-46 Resolutions Will Be Submitted,As Committed Previously

Forwards Response to 980202 RAI Re Summary Rept on Resolution of USI A-46, Seismic Qualification of Equipment in Operating Npps. Confirmatory self-assessment of USI A-46 Resolutions Will Be Submitted,As Committed Previously

ML20217J006
Person / Time
Site:	Crystal River
Issue date:	03/30/1998
From:	Rencheck M FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR 3F0398-16, 3F398-16, TAC-M69440, NUDOCS 9804060145
Download: ML20217J006 (15)

Text

.

~

Fl2rida i Power EE5!8"  !

M "* O *=. m .n

{

l March 30,1998 3F0398-16 U.S. Nuclear Regulatory Commission i Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Request for Additional Information Regarding Summary Report on the Verification of Seismic Adequacy of Mechanical and Electrical Equipment Dated December 31,1995 (TAC M69440)

References:

1. NRC to FPC letter dated February 2,1998 (3N0298-01)

2. FPC to NRC letter dated December 16,1997 (3F1297-24)

3. FPC to NRC letter dated August 15,1994 (3F0894-02)

Dear Sir:

Reference 1 provided a request for additional information (RAI) regarding Florida Power Corporation's (FPC) summary report on the resolution of Unresolved Safety Issue (USI) A-46, 4

Seismic Qualification of Equipment in Operating Nuclear Power Plants. Attachment B to this letter  ;

provides FPC's response to the RAI which identified four areas of interest.

In Reference 2, FPC documented two exceptions to the methods used to identify the safe shutdown paths and equipment in accordance with FPC's Plant Specific Procedure for resolution of USI A-

46. In Reference 3, FPC revised those exceptions. Attachment C to this submittal provides further <

clarification to those statements. Also in Reference 2, FPC committed to perform a confirmatory self-assessment of USI A-46 activities for Crystal River Unit 3 (CR-3). This effort will assess the i

.. completeness of the resolution of USI A-46 issues and will be completed prior to startup from Refueling Outage 11 (fall of 1999). As committed previously, FPC will forward the results of the ,

self-assessment to the NRC.

If you have any questions regarding this submittal, please contact Ms. Sherry Bernhoft, Manager, Nuclear Licensing at (352) 563-4566.

Sincerely,  ;

nh Director, Nuclear Engineering and Projects m n, MWR:aef 9804060145 980330 DR ADOCK 0500 2 Attachments p xc: Regional Administrator, Region 11 Senior Resident Inspector NRR Project Manager CRYSTAL RIVEP. ENERGY COMPLEX: 16750 W. Power Line Street

• Crystal River, Florida 344284708 + (362)7964488 A Florida Progress Company

i a

i FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 J DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 i FPC LETTER NUMBER 3F0398-16 I

ATTACHMENT A 1

l i

LIST OF COMMITMENTS j

~

f

..U.S. Nuclear Regulatory Commission - Attachment A' 3F0398 Page 1 of I o

List of Regulatory Commitments -

The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the. submittal represent intended or planned actions by Florida Power Corporation. They are described for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing of any ques,tions regarding this

- document or any associated regulatory commitments.

' COMMITMENT DUE DATE

)

FPC committed to perform a confirmatory self-assessment of USI A-46 November 1,1999 1 activities for Crystal River Unit 3 (CR-3). The effort will assess the completeness of the resolution of USI A-46 issues. 'As committed l

previously, FPC will forward the results of the audit to the NRC. Also as part of that self-assessment, FPC will consider the need for a more detailed emergency lighting analysis along with a detailed review of - -1 local operator actions to identify those that may be affected by potentially adverse environmental conditions resulting from the seismic event.- In addition, the assessment will include a validation of a previous review of the " bad actor" relay list.

FPC will evaluate the advantages of returning EFV-36 to a normally June 1,1998 open status to eliminate this potential single failure. The results of the j evaluation will be submitted to the NRC. '

I

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 FPC LETTER NUMBER 3F0398-16 ATTACHMENT B RESPONSE TO NRC REQUEST FOR ADDITIONALINFORMATION

~

. U.S. Nuclear Regulatory Commission Attachment B.

3F0398-16 Page 1 of 8 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION I l

, .NRC RM Quesdon 1:

Page 2 ofhe December 31,1995 submittal states in part that 17 " bad actor relays" uere enluated and 15 of hose were detennined to only genemte alarms uhidt wuld subsequently clear and therefore contact chatter muld not be considered a pmblem. ihe remaining tw relays uvuld only pcifonn an alannfunction and do not impact the operation of the diesels.

Assuming he alanns associated wik these 17 relays are expected to annunciate during the seismic b event, muld the operators have to respond to those annunciators and review the annunciator

[

response pmcedures associated with themforpotential action? How wuld those additional actions impact the operators ability to implement he Nonnal, Abnonnal, and Emergency'Opemting Procedures required to place the reactor in a safe shutdoun condition ?

FPC Response to RAI Question 1: l

, As a point of clarification, FPC performed a review of the " bad actor relays" associated with safe  !

I shutdown equipment which resulted in additional relays being identified and elimination of relays j not associated with safe shutdown equipment. As 'a result, the list has expanded to 74 relays, which .i actuate 71 different alarms. A validation of this most recent effort will be included within the -

o scope of FPC's confirmatory self-assessment. l l

I l

Operator response to annunciator alarms is dependent on plant conditions. If no plant transient is -  ;

l in progress, operators respond to annunciator alarms in accordance with annunciator response L . procedures. If a plant transient is in progress, the applicable Emergency Operating Procedure

- (EOP) or Abnormal Procedure (AP) takes precedence over the annunciator response procedure.

During response to a seismic event that does not result in reactor trip, operator ' actions are dictated by Operating Procedures (ops) and APs. AP-%1, " Earthquake," specifically addresses necessary l

. actions that may be required after a strong motion event. AP-%1 contains a listing of the 71 L spurious alarms that may be received, thus alerting the operator of the need to verify the alarm

through other indications. If the event does not result in the need for a reactor trip, operator actions to place the reactor in a safe shutdown condition would not be affected by this spurious alarming condition because a strong motion event is not a continuing occurrence. As such, the

~ spurious ~ alarming condition would only occur for a short duration (15 to 30 seconds) unless the device was a seal-in type relay. None of the subject relays have been identified as a seal-in type relay. If the alarm was the result of relay chatter, when the operator acknowledges the alarm the annunciator window will go dark. This will occur before operator actions are taken to place the reactor in a safe shutdown condition.

l

o ,

U.S. Nuclear Regulatory Commission Attachment B 3F0398-16 Page 2 of 8

• During normal operating conditions, ' thy operators are required to announce and respond to annunciator alarms in accordance with the associated Alarm Response (AR) procedure. However, l during transient conditions, in which multiple alarms occur while operator action and attention to L > main control board indications _ is required, normal alarm announcing and response may be l suspended. The method of performir.g the immediate actions necessary to place the reactor in a i safe shutdown condition are prescribed in Administrative Instruction (AI) 505, " Conduct of Operations During-Abnormal and Emergency Events." In a seismic event resulting in a reactor

!-  : trip, the operator is required to perform and verify the reactor trip immediate actions in EOP-02, i L Vital Systems Status Verification." One operator is required to perform the immediate actions, L while the second operator is monitoring plant instrumentation and alarms for the existence of L symptoms that would require entry into higher priority procedures _ such as loss of adequate -

L subcooling, upsets in primary to secondary heat transfer, or steam generator tube rupture. After initial completion of the immediate. actions, the Nuclear Shift Supervisor (NSS) directs re-u performance of these actions.' After completing the second performance of the reactor trip l immediate actions, the operating crew is required to pause to confirm there are no symptoms of

! more serious conditions prior to proceeding with the remaining portion of the reactor trip EOP.

Thus, because the procedure associated with the reactor trip has precedence over the annunciator response procedures, the occurrence of spurious alarms.during a seismic event would not challenge the operators' ability to assess plant stability or maintain safe shutdown.

N --NRC RM Onesdon 2:

Items 4.4.11 and 4.4.12 address lighting requiremenu and state, in part, that due local battery ponered light unit will remain in place and function as required qfter dse ewnt and that l supplemental light will be provided by battery powered and plug-in uniu connected to a diesel- l baded outlet.

l Describe the analysis performed to verify that local actions whids must be taken by operators as a result of the ewnt are in areas which are covered by either local battery operated lighu or in close pruimity to the diesel-backed unia. For those actions uhids require either diese!-boded or battery ponered lighting, is the equipment permanently staged in the areas required to be axessed by the l J operators? For those units uhids must be fed from the diesel-backed 'unin,, has the loading i associated with thesepieces of equipment on dse diesel been considered?

FPC Response to RAI Question 2:  ;

p:

i . .

H There is currently no specific analysis performed to verify that the areas in which local actions that ,

must be taken by operators (as a result of a seismic event) are covered by emergency lighting.' This is because permanently installed emergency lighting is not the primary lighting relied upon for performance of operator actions.' Plant operators rely on hand held lanterns and flashlights for the performance of manual actions required in the abnormal and emergency operating procedures.

~ Plant operators are required to carry flashlights with them as they make their rounds in the plant.

I 4

U.S. Nucler.r Regulatory Commission Attachment B

- 3F0398-16l Page 3 of 8 p

^

Additionally, eleven EOP Tool Boxes containing a six volt lantern are located around the plant.

I Six of these boxes also contain replacement flashlight batteries. These battery powered devices are

! used to aid in the performance of the required manual actions in the plant. The operators' ability to

! perform these actions under expected post-accident conditions has been verified as pan of the EOP validation process.

Although not relied upon, the load from the plant emergency' lighting'has been included in the emergency diesel generator loading calculations. FPC has committed to perform a confirmatory l- self-assessment of USI A-46 activities to assess the completeness of the resolution of A-46 issues.

As pan of that self-assessment, FPC will consider the need for a more detailed emergency lighting L -- analysis..

V l

[. . NRC RM Osession 3:

l Section 4.6, Opemtions Department Review of the Safe Shutdown Equipment List (SSEL), states thatplant operations representatives hme venfied that all equipment required in the opemting and emergency proceduresfor the selected safe shutdown path are included on he SSEL, and that these procedures are adequatefor plant response to the seismic ewnt. Additionally, in Section 4.6.1, i

Conclusion of Operations Deparonent Review, states that uhile some operator actions are required, adequate stagng, direction and time are available to acconplish the required actions.

Describe what reviews were perfonned to determine if any local operator actions required to safely shutdown the reactor could be afected by potentially adverse environmental conditions (such as loss of lighting, excessiw heat or humidity, or in-plant barriers) resulting from the seismic ewnt.

Describe how stagng uns evaluated and describe the resiews which were conducted to ensure operators had nd="mte time and resources to respond to such ewnts.

l FPC Response to RAI Question 3:-

No specific review has been performed to determine if any local operator actions required to safely - 1 shutdown the reactor could be affected by potentially adverse environmental conditions resulting from the seismic event. However, FPC has committed to perform a confirmatory self-assessment a of USI A-46 activities to assess the' completeness of the resolution of A-46 issues. As pan of that -l L - self-assessment, FPC will consider the need for a more detailed review of operator actions to L identify potential problems following a seismic event.

p

, A review has been performed to identify operator actions that were credited for an earthquake. As l described below, all shon term actions required to assure the plant is safely shutdown are  ;

performed from the Control Room. l Potential environmental challenges include toxic gases (sulfur dioxide and chlorine) which are stored at the coal plants. Should a release occur because of the seismic event, intrusion into the l

f

E .

~

~

U.S. Nuclear Regulatory Commission ~~ Attachment B 3F0398-16 Page 4 of 8

' plant is expected to be minimal because the plant is surrounded by a 20 foot high berm and both of

. the toxic gases stored on site are heavier than air. If the seismic event results in a loss of off-site power, the fans circulating air in the Turbine Building, Auxiliary Building, and Intermediate

- Building will stop since they are not backed up by the emergency diesel generator. This will serve to further limit gas intrusion into the buildings. - Additionally, wind tunnel testing has shown that-the effects of a gas release are short lived compared to the time available b-fore manual actions are required in the plant. As a result, little gas intrusion into the buildings is expected.

r Should a steam release occur in the Turbine or Intermediate Building due to the failure of non-

seismically designed piping, the steam release can be terminated remotely from the Control Room.

The longer term: actions required to be performed in the plant can be adequately planned and -

-: executed in the time available. These actions and the time available to perform them are described

!- below.

The Control Complex (including the Control Room), Auxiliary Building, Intermediate Building, Emergency Diesel Generator Building, and Reactor Building are seismic Class I structures.

Potential problems that could be created after a safe shutdown earthquake (SSE) as a result of

falling or failure of non-seismic components are therefore judged not to be significant in affecting l

any operator actions that could be necessary in these areas. Similarly, operator actions in the l Turbine Building is not expected to present a significant challenge or hazard with regard to an operator's ability to access plant equipment as required by procedure. Earthquake experience has shown that typical industrial grade equipment and structures are inherently- rugged'and not y susceptible to damage which would result in the structures or equipment inhibiting operator access b '

for a 0.lg level SSE. Additionally, plant areas that may require operator action are accessible via

' multiple pathways. Potential ingress / egress blockage in any one pathway would not prevent alternate access to plant areas.

Assuming the effects of the strong motion event result in a loss of off-site power, the' reactor is L placed in a safe shutdown condition using EOP-02, " Vital System Status Verification," and any

other EOPs and APs for which entry conditions may be met.' The systems and equipment required for achieving and maintaining the safe shutdown are identified as those which provide the following

. functions:

1 Reactor Reactivity Control l Reactor Coolant Pressure Control Reactor Coolant Inventory Control l . Decay Heat Removal

l. The following discussion addresses each of the above functions as related to a natural circulation j cooldown following a loss of off-site power (LOOP). While no intervening barriers resulting from j the strong motion event are assumed, plant procedures do contain additional compensatory actions  !

that will be discussed. 1 l .

I I^

~

[-

U.S. Nuclear Regulatory Commission 3F0398-16 Attachment B Page 5 of 8

' REACTOR REACTIVITY CONTROL Upon entry into EOP-02, reactor shutdown is achieved through control rod insertion. Due to the L physical control rod insertion limits required by technical- specifications, immediate reactor shutdown is assured upon the initiation of a reactor trip signal. Action to initiate a reactor trip would occur automatically through signals generated by the Reactor Protection System. However,

should a manual reactor trip be required, this action is accomplished by depressing a single button in the Control Room. As such, local actions are not required to effect a reactor shutdown.

EOP-09, " Natural Circulation Cooldown," requires the Reactor Coolant System (RCS) to be adequately borated to ensure adequate shutdown margin is maintained during cooldown. The boric acid storage tanks and the borated water storage tank are available to accomplish this function. If the concentrated boric acid storage tanks are used as the boration source, a local action to restore Instrument Air System operation may be required. This action is directed by AP-470, " loss of Instrument Air," which is referenced from the reactor trip EOP. This action is not required until primary system cooldown is commenced, and involves one operator manually starting a diesel driven air compressor. Recent procedur validations of field actions have shown that once the operator is dispatched, the time required Or completing the action to start the air compressor is less than 5 minutes. An evaluation done to dumonstrate compliance with Appendix R concluded that adequate condensate sources are ava.ilable to delay initiation of RCS cooldown for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time period. A summary of this ev&ation was submitted to the NRC in a letter dated December 10, 1997 (3F1297-04). This 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provides ample time for planning the restoration of the Instrument Air System ope ration.

The flow path from the borated water storage tank contains motor operated, Engineered Safeguards -

(ES) powered valves which are operated from the Control Room. Therefore, if the borated water storage tank is used as the boration source, no field actions are required.

REACTOR COOLANT PRESSURE CONTROL The normal methods of RCS pressure control following a LOOP condition are through use of high pressure auxiliary pressurizer spray to reduce pressure, and the pressurizer heaters to increase or maintair. pressure. Both of these control methods require local operator action for alignment or restoration. Since elevated RCS pressure is not expected fn>m a reactor trip, and since the power

' operated relief valve (PORV) is available for RCS pressure reduction, restoration of pressurizer heaters is considered more critical. Action to restore pressurizer heaters is directed by AP-770,

" Emergency Diesel Generator Actuation," which is referenced from EOP-02. Recent procedure validations of field actions have shown that once the operator is dispatched, the time to complete all of the necessary local actions to restore pressurizer heater operation is less than 10 minutes.

L During the performance of this procedure, the local actions are combined with Control Room I

actions that involve selecting various switches on the main control board. The rate of heat loss from the pressurizer does not result in a rapid reduction in RCS pressure. Thus, adequate time for ,

I' completing the procedure for restoration of pressurizer heaters is available. An additional l

c ,

I

- , U.S. Nuclear Regulatory Commission Attachment B 3F0398-16 Page 6 of 8 i

J L ' l l contmgency action for RCS pressure control provided in the natural circulation cooldown EOP is to

. establish solid RCS pressure control. This can be accomplished from the Control Room without l ~ local operator action in the plant.

Action to reduce RCS pressure is not required until primary system cooldown is commenced.

Recent procedure validations of field actions have shown that once the operator is dispatched, the

time required for completing the action to align high pressure auxiliary spray is less than.20 i minutes. As stated previously, an evaluation done to demonstmte compliance.with 10'CFR 50, Appendix R concluded that adequate condensate sources are available to delay initiation of RCS 1 cooldown for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time period, thus providing ample time for planning the alignment of high pressure auxiliary spray.

)

REACTOR COOLANT INVENTORY CONTROL The primary method for RCS inventory control is accomplished using the Makeup and Purification System. The same local action previously discussed may be required to restore Instrument Air System operation to enable the normal control method of the system. As an additional contingency action, the reactor trip EOP contains guidance to utilize flow from the borated water storage tank,.

through a high pressure injection valve if normal makeup flow is not maintaining pressurizer level.

The flow path from the borated water storage tank contains motor operated, ES powered valves which are operated from the Control Room. Therefore, if the borated water storage tank is used as the makeup source, no field actions are required.

L DECAY HEAT REMOVAL i I

L The establishment of decay heat removal post-trip does not require local actions. . During a LOOP condition, decay heat removal is accomplished through automatic initiation of the Emergency .!

- Feedwater Initiation and Control (EFIC) System,' while steaming through the main steam safety  !

valves (MSSVs). 'lhe same local action previously discussed'may be required to restore Instrument '

- Air System operation, enabling normal control of the atmospheric dump valves (ADVs). Operation  ;

of the ADVs is required when primary system cooldown is initiated. As previously discussed,  !

adequate condensate volume is available, thus providing ample time for planning the restoration of {

u iInstrument Air System operation. i As'an additional contingency action, local. manual operation of the ADVs can be performed.

Direction for' establishing manual control of the ADVs is contained in OP-608, "OTSGs and Main Steam Systems," as well.as posted locally at the valve. - Personnel access' to the Intermediate Building may not be possible based on calculated atmospheric temperature. However, opening the doors leading to the Turbine Building results in a temperature reduction such that intermittent

, 1 access to. the Intermediate Building is possible to adjust the ADVs. AP-%1, " Earthquake,"

I contains the action to nen the Intermediate Building doors.

I J f

[ .

'< ,U.S. Nuclear Regulatory Commission

!- Attachment B l 3F0398-16 Page 7 of 8 Continued decay heat removal during a IDOP condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> requires use of condensate sources in excess of the emergency feedwater tank. The additional condensate sources available are

! the condensate storage tank and the condenser hotwell. EOP-09, " Natural Circulation Cooldown,"

contains instructions to perform the necessary local operator actions to align these sources for

suction. to the emergency feedwater pumps. , AP-%1, contains guidance;to perform plant assessments to assure that the necessary components to perform emergency feedwater pump suction transfer, can be accessed. If this equipment cannot be accessed, guidance is provided to perform j the necessary action to allow access. The emergency feedwater tank contains adequate volume to I remove decay heat for approximately_18 hours,' thus providing adequate time te perform the

' necessary actions to allow access to these components.

l' NRC RM Question 4:

' 1 i

Table 5.4,' Outliers, of the ewluation report, identify several contml room structurc uhich could l impact the opemtor's ability to respond to the seismic ewnt, including the potentialfor the main  !

contml mom ceiling to be an interaction source, several non-bolted cabinets, and several non-restminedpieces of equipment (i.e., computer keyboard and stand). Haw each of these potential  ;

sources ofintemctions been enluated and have thejinal resolutions to eads been inplemented? .

FPC Response to RAI Question 4:

FPC has identified and evaluated each source of interaction concern noted in. Table 5.4. The

. Seismic Review Team (SRT) was cognizant of issues related to interaction sources inside the Control Room. This resulted in several pieas of plant equipment in the Control Room being declared outliers. . The following is a list of the equipment in the Control Room and current status:

L .1. Control Room Ceiling: j I

The Control Room ceiling was identified as being a potential interaction problem and an outlier.-

in the original walkdown. During the 1997 outage, FPC completely replaced the Control 3 Room ceiling with a ceiling system specifically designed and installed to be seismically rugged, to provide better lighting, and to lower sound levels. The ceiling replacement project was

. completed prior to restart. No further action is required to resolve this issue.

L 2. ' Non Nuclear Instrumentation (NNI) Cabinets and Integrated Control System (ICS) Cabinets:  !

l The seismic review identified several NNI and ICS cabinets that were not bolted together. The L cabinets were anchored and therefore will not create any interaction effect that would cause a l

barrier to operator actions. . The fact that the cabinets were not bolted together leads to an  !

interaction concern only in terms of relay functionality. The cabinets do not present a falling I'

- hazard; therefore, no immediate further action is required. However, the cabinets are being

I' , y

~ '

U.S. Nuclear Regulatory Commission Attachment B 3F03%16 Page 8 of 8 mddified to resolve the relay functionality concern. This is being done under an existing commitment (Reference 2).

3. Event Recorder (ER) Cabinets:

Like the NNI cabinets, the ER cabinets were conservatively added to the SSEL because they might contain essential relays. The SRT could not determine adequate anchorage by field inspection. The internal wiring of these cabinets was very congested and prevented a visual inspection of all the anchorage. Thus, they were all declared outliers pending further review.

FPC has completed a calculation (S98-0541) that shows the cabinets will not fall over. . This -

calculation conservatively assumed the cabinets were not anchored. The cabinets may rock

slightly, but would not fall over or hinder operator movement in the Control Room. These cabinets are all bolted together. Since these cabinets do not contain essential relays, and the conservative calculation shows the cabinets will not fall, no further action is required for.these cabinets.

L 4. Engineered Safeguards (ES) Actuation Relay Cabinets:

l.

Four ES relay cabinets were identified as outliers because there was an interaction concern with nearby tool boxes. The SRT postulated the tool boxes could fall and impact the cabinets. This has been resolved by moving the tool boxes to a less sensitive area. There is no longer an  !

interaction affect with this equipment. There were no operator concerns for these outliers. . The cabinets and tool boxes are located in a corner of the Control Room that is outside'of any L operator traffic area.

A missing latch on a panel inside one of the cabinets has been replaced. 'Ihere is no further l action required to address these cabinets.

5. Nuclear Instrumentation and Protection (NI&P) Cabinet D2:

p L

This cabinet was identified as an outlier because a book stand was located next to the cabinet during the seismic review walkdown. This is a generic interaction concern that was brought to the attention of operations personnel. The' book stand has been replaced with a more stable stand and has been relocated to an area where it will not impact essential equipment. Falling of the book stand will not affect any operator traffic. There is no further action required to address this issue.-

~

In summary, FPC has evaluated sources of seismic interaction concern in the Control Room. For

. each potential interaction source, appropriate actions have. been taken. There is'no further action E

- required beyond existing commitments.

4

. FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302-LICENSE NUMBER DPR-72 FPC LETTFsR NUMBER 3F0398-16 ATTACHMENT C SINGLE FAILURE CONSEQUENCFE ,

1 FOLLOWING A SEISMIC EVENT i

l

~

' U.S. Nucicar Regulatory Commission Attachment C 3F0398-16 Page 1 of 2 SINGLE FAILURE CONSEQUENCES FOLLOWING A SEISMIC EVENT Steam Relief Capacity The Plant Specific Procedure (PSP) for Seismic Verification of Nuclear Plant Equipment defines safe shutdown as bringing the plant to, and maintaining it in, a hot shutdown condition during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a safe shutdown earthquake (SSE). For Crystal River 3 Unit (CR-3), hot shutdown is defined as an average Reactor Coolant System (RCS) temperature less than 280 F.

~

The PSP also states that systems selected for accomplishing safe shutdown should not be dependent

. upon a single item of equipment whose failure, either due to seismic loads or random failure, would preclude safe shutdown. At least one practical alternative should be available for accomplishing safe shutdown which is not dependent on that item of equipment.

In a submittal dated August 15,1994 (3F0894-02), FPC provided the discussion of the methodology for achieving and maintaining hot standby following a design basis seismic event at CR-3. 'Ihe FPC A-46 program scope includes the systems and corresponding equipment necessary ;

to ensure that hot standby can be achieved and maintained for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an SSE. It was believed at that time that it was not possible to cooldown to a hot shutdown condition within 72 -

hours given a loss of off-site power (LOOP). FPC stated at that time that, for this condition, an additional 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> was required to cool the RCS down to the 280 F. This was due to the limited  ;

capability to relieve steam through the atmospheric dump valves.

This exception was formally accepted by the NRC in the Safety Evaluation Report on the PSP l dated May 2,1996 (3N0596-04). It stated, "The staff concluded that the licensee's approach to 1 achieve and maintain hot standby for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an SSE is acceptable if the licensee u confirms that the equiprnent necessary to assure core decay heat removal for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in both of the safe shutdown paths is seismically adequate." The seismic adequacy of that' equipment was confirmed through the plant walkdowns performed in accordance with the PSP. J

~In a more recent submittal dated December 16,1997 (3F1297-24), FPC modified its earlier position by stating that recently performed calculations of the plant response to a natural circulation cooldown using a more sophisticated model and 'more realistic assumptions demonstrated that the-plant can be cooled to less than 200 F in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This model includes a more realistic decay heat model and takes credit for the heat' remover through operation of the turbine driven emergency feedwater pump, in addition to the two atmospheric dump valves.

From this, it was concluded that there were no identified exceptions to the methodology used to identify safe shutdown paths and components in accordance with the PSP, However, there are two requirements of the PSP at issue, both of which should have been explicitly addressed in the previous correspondence.

r;

, ]

.- U.S. Nuclear Regulatory Commission Attachment C .

3F0398-16. Page 2 of 2

1. There is a PSP requirement _ to demonstrate the capability to perform a cooldown to hot shutdown in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Based on the results of the more recent calculation, FPC takes no exception to this requirement.

( 2. There is an additional PSP requirement that the cooldown to hot shutdown should' not be L._ dependent upon a single item of equipment whose failure, either due to seismic loads or random failure, would preclude safe shutdown. FPC must still take exception to that requirement. FPC's calculations show that failure of either of the atmospheric dump valves or

. the turbine driven emergency feedwater pump would prevent the cooldown from being.

performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. These three' steam release paths are all necessary to maintain the L required cooldown rate. Failure of a steam release path does not affect the ability to remove

! decay heat and maintain hot standby for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

l

' Availability of Condensate Sources L Continued decay heat removal during a LOOP condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, as required by the PSP, l requires use of condensate sources in addition to the emergency feedwater tank. The additional L condensate sources available are the condensate storage tank and the condenser hotwell. These sources have been verified to be seismically rugged in accordance with the PSP.

L .

. i

!- It has recently been discovered, however, that one 'of these sources may not be available when a

-single failure is considered. There is a single line between the mndenser hotwell and the suction of .

L both emergency feedwater pumps which contains a manual valve which was locked open when the Safe Shutdown Equipment List (SSEL) was prepared. Subsequent to that, the valve was closed for j

. operational considerations.- This valve must be opened to transfer emergency feedwater pump suction to the condenser hotwell. Failure of this valve in the closed position eliminates' access to about 150,000 gallons of condensate grade water. The remaining volume is sufficient to support decay heat removal and plant cooldown for approximately 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />. Failure of the valve in the closed position following a seismic event is considered a low probability event since the valve is-l

-inherently rugged per the CR-3 Plant Specific Procedure, and the failure of manual valves to open o is also low (3.73X104per demand). FPC will evaluate the advantages of returning this valve to a -j

. normally open status to climinate this potential single failure. This evaluation will be completed ' 1 l and the NRC will be notified of the results by June 1,1998. This is considered a timely response 4

since the probability of a SSE in the intervening time is less than 4.0X10 . In addition, EM-225,

" Duties of the Technical Support Center Accident Assessment Team," contains guidance to begin preparations for a backup emergency feedwater source in case the existing source degrades or fails.

l l

l La