ML20137L786
| ML20137L786 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 10/25/1994 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20137K821 | List:
|
| References | |
| FOIA-96-485 JPN-PSL-SEMP-94, JPN-PSL-SEMP-94-076, JPN-PSL-SEMP-94-76, NUDOCS 9704070315 | |
| Download: ML20137L786 (25) | |
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FLORIDA POWER AND LIGHT
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SAFETY EVALUATION l
1 INCREASE OF ENGINEERED SAFEGUARDS SUCTION PIPING DESIGN PRESSURE l
St. Lucie Unit 1 6
JPN-PSL-SEMP-94-076 Revision 0 SAFETY RELATED 1
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JPN-PSCSEMP-94-076 I
Revision 0 l.
Page 2 of 12 REVIEW AND APPROVAL RECORD i:
i PLANT St.12ncle UNIT 1
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I TITLE Incr== of Fneineered Safernards Suction Ploine Deslen Pressure j..
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l LEAD DISCIPLINE Mechanical i
i ENGINEERING ORGANIZATION St. Lucie Production Eneineerine Groun 1
i REVIEW / APPROVAL:
INTEtFACE TYPE l
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OTHER INTERFACES Fonn 24. Rav 6/M
I-JPN-PSL-SEMP-94-076 Revision 0 Page 3 of 12 j
l Table of Contents i
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., 3 Cover Sheet 1
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Interface Approval Record 2
Table of Contents 3
1.0 Abstract 4
j 2.0 Purpose / Scope 5
i 3.0 Licensing Requirements 6
4.0 Ana[ysis of'iffects on Safety 6
l 5.0 Failure Modes and Effects Analysis 1 g
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6.0 Plarit Restrictions g
l 7.0 Effect on Technical Specifications g
8.0 Unreviewec Safety Question Determination 9
i 9.0 Conclusi.,as 11 i
10.0 Verification Summary 11 11.0 References 12
' Attachment 1 Minimum Wall Calculation 2pgs i
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1.0 ABSTRACT After successful performance of a motor operated valve' differential pressure'testwyeijpdfed%y '.x t' r j
NRC Generic Letter 89-10, water was observed on the floor of the pipe tunnel in the Reactor Auxiliary Building. The source of the water was determined to be from relief valve, I-SR i 1 A, located on the 1 A Engineered Safeguards suction piping.
i An examination of the alignment used for the test revealed a flow path from the discharge of the i
1B Containment Spray pump to the suction of. the 1A Containment Spray pump through a i
common header in the Sodium Hydroxide (NaOH)' Spray Additive systeene With-the 1Rm Containment Spray pump operating am' all theta ECCS pumps secured, theWtfdin(suction " -
piping was pressurized through the Spray Additive system common header. As part of the j
investigation of the event, the IB Low Pressure Safety Injection pump was aligned to the B l
Containment Spray header and discharged through the B Shutdown Cooling heat exchanger. The maximum pressure observed at the 1 A I.aw Pressure Safety Injection pump discharge header was 80 psig. Relief valve I-SR-07 ' f W determined to discharge. Plant personnel documented the event on St. Lucie Action Request (STAR) 1-94100259, Reference 11.1, which was assigned i
to Nuclear Engineering for disposition.
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An Initial Assessment of Operability was performed to respond tc the concerns addressed by the l
STAR. A calculation determined that the components whose design pressure had been exceeded j
were in fact capable of withstanding considerably higher pressures. The suction piping and components, therefore, did not suffer any damage as a result of the event. However,.with _
rcgard to the issue of the system design, it was concluded that a design basis scenario exists i
which could result in lifting the relief valve; a containment spray actuation signal (CSAS) and i
a loss of off-site power (LOOP) coincident with one Emergency Diesel Generator (EDG) failing to operate. The relief valve could potentially open and release containment sump inventory in excess of the Engineered Safeguards equipment external leakage rate of 2 liter per hour, FSAR i
15.4.1.7 and 15.4.1.8, Reference 11.2. This could result in a condition outside of the design basis of the Engineered Safeguards systems. Based on the Initial Assessment of Operability, j
plant management made a one hour non-emergency notification to the Nuclear Regulatory Commission in accordance with 10 CFR 50.72.
This safety evaluation was prepared to evaluate the acceptability of higher pressures in the Engineered Safeguards suction piping in order to disable relief valves I-SR-07-1 A and I-SR !
1B for this cycle while above Mode 4. This interim measure is being implemented to prechule the possibiliy of an unwanted release. The proposed change does not require a permane.nt i
change to the facility. The safety evaluation addresses the acceptanility of implementing tha proposed change during operation. This safety evaluation invob :s Engined Safeguards j
systems and is therefore classified as safety related.
l The proposed change has been reviewed to determine whether an unreviewed safety question j
exists or if the plant Technical Specifications are affected. Based on the evaluation herein, it has been determined that an unreviewed safety question does not exist and the plant Technical i
Specifications are not affected. Therefore, prior notification of the NRC is not required.
h, JPN-PSL-SEMP-94476 Revision 0 Page 5 of 12
2.0 DESCRIPTION
ANP FURPOSE.
P After sn~=ful performance of a motor operated valmdifferentisipessene test asaquimd by P A NRC Generic I.stter 89-10, water was observed on the flooe of the pipe.tunnelin the Reactor Auxiliary Building. The source of the water was determined to be from relief valve, I-SR 1 A, located on the 1A Engineered Safeguards suction piping.
An examination of the alignment used for the test revealed a flow path from the discharge of the IB Containment Spray pump to the suction of the 1A Containment Spray pump through a common header in the Sodium Hydroxide (NaOH) Spsay Additive system. With the IB.
Containment Spray pump operating and all the A ECCS pumps secured, the A train suction l
piping was pressurized through the Spray Additive system common header. As part of the investigation of the event, the IB I.ow Pressure Safety Injection pump was aligned to the B Containment Spray header and discharged through the B Shutdown Cooling heat exchanger. The maximum pressure observed at the 1 A 1.ow Pressure Safety Injection pump discharge header was L
80 psig. Relief valve I-SR-07-1 A was determined to discharge. Plant personnel documented l
the event on St. Lucie Action Request (STAR) 1-94100259, Reference 11.1, which was assigned.
to Nuclear Engineering for disposition.
An Initial Assessment of Operability was performed to respond to the coxerns addressed by the l
STAR. A calculation determined that the components whose design pressum had been exceeded
[
were in fact capable of withstanding considerably higher pressures. 'Ihe suction piping and components, therefore, did not suffer any damage as a result of the_ event._.However,.with J
regard to the issue of the system design, it was concluded that 7. d:dgn basis scenario exists i
which could result in lifting the relief valve; a containment spriy actuatior, signal (CSAS) and a loss of off-site power (LOOP) coincident with one Emergency Diesel Generator (EDG) failing to operate. The relief valve could potentially open and release containment sump inventory in l
excess of the Engineered Safeguards equipment external leakage rate of 2 liters per hour, FSAR i
l 15.4.1.7 and 15.4.1.8, Reference 11.2. This could result in a condition outside of the design basis of the Engineered Safeguards systems. Based on the initial Assessment of Operability, plant management made a one hour non-emergency notification to the Nuclear Regulatory Commission in accordance with 10 CFR 50.72.
This safety evaluation was prepared to evaluate the acceptability of higher pressures in the Engineered Safeguards suction piping in order to disable relief valves I-SR-07-1A and I-SR IB for this cycle while above Mode 4. This interim measure is being implemented to preclude the possibility of an unwanted release. The proposed change does not require a permanent l
change to the facility. The safety evaluation addresses the acceptability of implementing the proposed change during operation. This safety evaluation involves Engineered Safeguards systems and is therefore classified as safety related.
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Page 6 of 12 1
3.0 LICENSING REQUIREMEN'IE T "5 e.ro.
De ECCS suction pipingis designed irraccordance with1heig-i- -
of.USAS B31!7e1969,' -
l Class IL The code does not provide a specific requirement for relieving capacity in this portion of the system. The relief valves, I-SR-07-1A and I-SR-07-1B, are not specifically addressed in either the Final Safety Analysis Report or the plant Technical Specifications. Based on meeting minutes, Ebasco Services Inc., the architect / engineer, provided the relief valves because of a concern over the interface between the low pressure suction piping and the portion of suction j
piping used for Shutdown Cooling. Ebasco Services recommended installation of check valves and relief valves to protect against leakage across the motor operated valves or the failure to j
isolate this portion of the system prior to initiating Shutdown Cooling, Reference 11.6.
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4.0 ANALYSIS OF EFFECTS ON SAFETY A credible design basis scenario is pedaw that could open the relief valve. In this event, a l
containment spray actuation signal (CSAS) occurs, a loss of off-site power (LOOP) occurs and j
one Emergency Diesel Generator (EDG) fails to operate. The postulated accidents when these l
conditions exist are a large break loss of cooling accident (LOCA) with a LOOP, or a Main Steam Line Break (MSLB) inside containment with a LOOP. The event results in the actuation i
l of a Containment Spray pump with no ECCS pumps operstmg on the opposite train.
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In this scenario, it is expected that both Containment Spray flow control valves, I-FCV-07-1 A and I-FCV-07-1B would open These are air operated, fail cpen. valves.__nese valves open on loss of air and/or power. CSAS provides a signal to de-;.-pze the solenoid valves FSE 07-1 A l
and FSE-07-1B which open I-FCV-07-1 A and I-FCV.07-1B respectively (References 11.8,11.9, l
11.10 and 11.11). This would provide a vent path to the containment. Based on the height of the containment spray ring header, the peak containment accident pressure and an eductor design flow rate of 128 gpm, Reference 11.15, the maximum expected pressure in the Engineered i
Safeguards suction piping would not exceed 150 psig. A review of the isometric drawings for I
the Engineered Safeguards suction piping shows that all flanges and fittings have, as a minimum, l
an ANSI rating of 150 lb. Since the pr.ak post accident temperature is 200* F downstream of p j
the Shutdown Cooling heat exchanger, the allowable pressure becomes 235 psig, Reference 1 L7.
! provides a bounsing calculation that demonstrates that the piping is acceptable for I
internal pressures of 250 psig.
l The valve packing, flanges and mechanicaljoints have not been pressure tested to the ANSI 150 i
Ibs. Class rating corresponding to the peak accident temperature (<200'F). However, they are i
installed and torqued to the ANSI 150 lb. Class rating. To further substantiate the design, a leak
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check should be performed to ensure the leak tightness of the valve packing, i
mechanical joints. The affected piping is the Engineered Safeguards suction pipmg in the i
Recirculation mode. Based on an RWT temperature of 100"F or less, the test pressure should no: exceed 250 psig. External leakage should not exceed the limits as specified in FSAR Table j
15.4.1-2.
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l' JPN-PSL-SEMP-94-076 Revision 0 Page 7 of 12 l:
4.0 ANALYSIS OF EFFECTS ON SMFETY (Contliised) 3 " '
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It is assumed a vent path is open WIhfoontainmen't'through the ifi,i.,lehiEK 'rinf
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spray header. This would result in slightly less flow being~ delivered to the active Containment j
Spray header. The containment spray pumps are tested quarterly in accordance with the plant's i
In-Service Test program, which is consistent with ASME B&PV code Section XI, Table IWP-3100-2.Section XI paragraph IWP-3230 requires that the pump maintain.;>0.90 differential i,
pressure of its reference value for operability. The reference value is nominally the pump l
performance curve, his was validated by revie.w of the latest surveillance for the 1 A and IB containment spray pumps.
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The pump and system resistance curves are provided in the calculation for the containment spray
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system flow and pressure dro'p (PSL-lEJM-90-050, Reference 11.16). Plotting a containment l
spray pump performance of 90% of nominal pump performance demonstrates that the pump is i
capable of injecting over 3100 gpm into containment at peak LOCA pressure. Since only 2700 I
gpm is assumed in the containment analysis, there is an excess of 400 gpm. As discussed previously, the design flow of spray additive system eductor is 128 gpm, which is much lower i
than the excess flow rate of 400 gpm available.
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nerefore, the design flow rate of 2700 gpm to the spray header will be assured even if some j
of the spray pump flow is diverted through the idle spray pump's header.
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Another scenario is postulated in which the vent _ path to the containment may_no.t be available.
Although it is highly unlikely, should an event occur which requires the initiation of the recirculation mode and Containment Spray was either not initiated or has been secured, the vent 1
path through either I-FCV-07-1 A or I-FCV-07-1B would not be available. Therefore, the use of Containment Spray to discharge through the Shutdown Cooling heat exchangers to provide i
j additional cooling while supplying the High Pressure Safety Injection pumps during recirculation
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could, pressurize the idle train suction piping to the outlet conditions of the Shutdown Cooling, l
heat exchang'er. During the recirculation mode, the e.xpected temperature and pressure are 200'F 1-i afnd 225 psig, respectively..,At-these conditions, ANSI 150 lbs. f1Egd ahd j
al'lowabledesign pressure of 235 psig. As previously demonstrated, the piping, valves and i
/ fittings is acceptable for internal pressures of 250 psig.
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/ The record stress analysis calculations (References 11.12) and stress isometeric drawings l
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(Reference 11.13) have been reviewed for the affected ECCS suction piping and it has been j
I determined that an increase of internal pressure up to 250 psig has no adverse effect on the 7
f piping system.
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The requirements for valves and packing in the Engineered Safeguards suction have been l
reviewed. Based on the St. Lucie Mechanical Maintenance packing program or the original equipment manufacturers' packing requirements, an internal pressure of 250 psig should not
\\ adversely affect valve packing. Operational requirements for valves are not affected by the
' increase in pressure.
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JPN-PSL-SEMP-94-076 i
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4.0 ANALYSIS OF EFFECTS O,N SAFETY.(Con't==ad) w. j i
ne valve outline diswing, Reference 11.14, provides an acceptable.means for disabling the i
relief valves. Note the blocking device should only be installed " hand tight". He relief valves
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should be bench tested and re-set, if required, during the upcoming refueling outage prior to i
returning to power operation. Based on the review of the design and a successful inservice leak j
test, disabling the relief valves is acceptable during normal operation. However, with Shutdown l
Cooling in operation, the same concerns exist as outlined by Ebasco Services. To preclude these concerns, the relief valve should be restored to service prior to entering Mode 4. He relief j
valve will still be capable of providing adequate protection in the Shutdown Coohng mode. The l
installed spring has a setting range of.55 to 67 psi which is adequate. The piping had been j
pressure tested to 76 psi.
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5.0 FAILURE MODES AND EFFECTS ANALYSIS
- I The proposed change described above does not adversely affect the Engineered Safeguards l
suction piping. Disabling the relief valves serves to reduce the possibility of the relief valves i
leakmg, actuating prematurely or failing to re-seat. The relief valves are no longer an active l
component and the failure modes are reduced. Additionally, FSAR Table 6.3.3, " Single Failure l
Analysis, High Pressure Safety Injection", Reference 11.2, was reviewed and it was determined j
that the relief valves did not affect the analyses. Therefore, no new failure modes have been i
introduced. Based on the above, no new failure modes are created.by proposed change, g
f 6.0 PLANT RESTRICTIONS 4
i The proposed change does not affect normal plant operations. Operations should be aware that the original intent of these relief valves was to protect against leakage across the motor operated valves or the failure to isolate this portion of the system prior to initiating Shutdown Cooling.
The relief valves are not addressed by any plant Technical Specification. This safety evaluation 1
i is valid for the remainder of the current fuel cycle. Prior to entering Mode 4, the relief valves, I SR-07-1 A and I-SR-07-1B, should be returned to their original design configuration, i
l 7.0 EFFECTS ON TECHNICAL SPECIFICATIONS The St. Lucie Unit 1 plant Technical Specifications, Reference 11.3, have been reviewed. The sections reviewed include, but are not limited to, sections 3/4.5, Emergency Core Cooling i
Systems and 3/4.6, Containment Systems. The relief valves, I-SR-07-1 A and I-SR-07-18, are l
not discussed within the plant Technical Specifications. There are no Technical Specifications i
bases adversely affected by the proposed change. The proposed change, therefore, does not have an adverse affect on plant Technical Specifications.
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JPN-PSL-SEMP-94-076 Revision 0 Page 9 of 12 8.0 UNREVIEWED SAFETY QUESTION DETERMINATION 4
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10 CFR 50.59 allows the holder of a license authorizing operation of a production or utilization facility to (I) Make changes in the facility as desenbed in the Safety Analysis Report, (II) Make changes in the procedures as described in the Safety Analysis Report, and (III) Conduct tests or experiments not described in the Safety Analysis Report, without prior Commission approval, unless the proposed change, test, or experiment involves a change in the Technical Specifications incorporated in the license or an Unreviewed Safety Question.
As defined in 10 CFR 50.59, an unreviewed safety question exists; (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety-previously evaluated in the Safety Analysis Report (SAR) may be increased; or (ii) if a possibility of an accident of malfunction of a different type than any pre dously evaluated in the SAR may be created; or (iii) if the margin of safety as defined in the basis for any Technical specification is reduced.
In 'accordance with 10 CFR 50.59, the following evaluation serves to determine whether this modification constitutes an unreviewed safety question or requires a change to the Technical Specifications:
8.1 Does the prooosed chanee increase the orobability of occurrence of an accident oreviousiv evaluated in the SAR?
The proposed change does not a'ffect any equipment whose malfunction is postulated in the SAR to initiate an accident or prevent an accident from occurring. As such, the I
probability of occurrence of an accident previously evaluated in the SAR has not increased.
8.2 Does the orooosed chance increase the consecuences of an accident oreviously evaluated in the SAR?
The proposed change does not affect the design function of any equipment designed to mitigate the consequences of an accident previously evaluated in the SAR. No new failure modes are being introduced and the design margin of equipment important to safety is not being decreased. As such, the proposed change does not increase the consequences of an accident previously evaluated in the SAR.
JPN-PSL-SEMP-94-076 Revision 0 Page 10 of 12 8.3 Does the orooosed channe inemw the otobability of an arenrrence of a malfnan+ian of eouinment imnortant to safety oreviousiv evalunted in the SAR?
'Ihe component affected by this change is the Engineered Safe. guards suction piping. 'Ihe proposed change maintains the quality level and the level of protection previously established for the Engineered Safeguards suction piping. The design allowable pressure rating of the piping and associated components is above the maximum sjstem pressure resulting from this change. Although the new higher pressure is above the original tested pressure, the design margin between allowable stresses and ultimate capacity is not being decreased. The proposed change, therefore, does not affect the pressure boundary integrity of the Engineered Safeguards suction piping. As such, the probability of an occurrence of a malfunction of equipment important to safety previously evaluated in the SAR is not increased by the proposed change.
8.4 Does the omoosed channe incran=* the con =anuences of a malfunction of couinmaat imoonant to safety oreviousiv evaluntad in the SAR?
The malfunction evaluated in the SAR is the complete or parnal failure of one train of Engineered Safeguards to perform its function. The proposed change does not in any way
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affect the ability of tne redundant train of Engineered Safeguards to perform its function 1
to inject and recirculate borated water. Therefore, the consequences of a malfunction of equipment important to safety previously evaluated in the SAR is not increased by the proposed change.
8.5 Does the orooosed chance create'the nossibility of an accident of a different tvoe than l
- g;v oreviousiv evaluated in the SAR?
l As discussed in Section 5.0, the proposed change does not introduce any new failure modes. The proposed change serves to reduce the possibility of a component failure.
l Therefore, the possibilitysof an accident of a different type than any previously evaluated m the SAR is not created by this proposed change.
8.6 Does the orooosed channe create the oossibility of a malfunction of equioment imoortant to safety of a different tvoe than any oreviousiv evaluated in the SAR?
l The proposed change does not interact spatially or functionally with any structure, system l
or component important to safety other than the Engineered Safeguards suction piping.
No new failure modes are created by the proposed change that can be postulated to cause a malfunction of equipment important to safety different than those previously analyzed j
in the SAR. Therefore, the possibility of a malfunction of equipment important to safe.ty which is of a different type than any previously evaluated in the SAR is not created by the proposed change.
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1 JPN-PSL-SEMP-94-076 Revision 0 i
Page 11 of 12 8.7 Does the oroW chane mduce the marein of safety as definad in the buin for any Techmcal Specification?
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r De Technical Specification requirements and Technical Specification Bases are not affected by the proposed change. The proposed change does not affect any plant Technical Specification requirement, as stated in Section 7.0. The proposed change maintains the quality and level of protection previously evaluated in the SAR. Therefore, the margin of safety as defined in the bases for any Technical Specification is not i
reduced by this proposed change.
9.0 CONCLUSION
AND RECOMMENDED ACTIONS The implementation of the proposed change has been reviewed against the reqturements of.10 CFR 50.59. ne disabling of the relief valves, I-SR-07-1 A and I-SR-07-1B, does not affect the plant Technical Specifications nor do they constitute an Unreviewed Safety Question. Based on the above, implementation of the proposed change without prior notification to'the NRC is I
acceptable. The following actions are recommended:
9.1 To further substantiate the design, a leak check shedd be performed to demonstate the leak tightness. The test pressure shall not exceed 250 psig based on a RWT water temperature of 100'F. He external leakage should not exceed the limits as stated in FSAR Table 15.4.1-2. Tests results should be reviewed and accepted.
9.2 The outline drawing of the relief valve identifies an acceptable means of disabling the relief valves. Note the blocking device should only be installed " hand tight".
The relief valves should be bench tested and re-set, if required, during the upcoming refueling outage prior to returning to power operation.
9.3 The relief valves should be returned to their original configuration prior to entering Mode 4.
10.0 VERIFICATION
SUMMARY
The scope of this verification was to review the inputs and references to determine if the results/ outputs were reasonable in comparison to the inputs. The method used for this verification consisted of ensuring applicable codes, references and regulatory requirements were addressed and properly reflected herein.
The verifier concurs with the safety related classification on this engineering evaluation.
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11.0 REFERENCES
11.1 St. Lucie Action Request Number 1-94100259, dated Oc:ober 21,.1994 -
11.2 St. Lucie Plant, Unit 1 Updated Final Safety Analysis Report, Amendment 13 11.3 St. Lucie Plant, Unit 1 Technical Specifications, Amendment 128 5~
11.4 St. Lucie Unit 1, 8770-G-088, Revision 27, Containment Spray and Refueling Water i
System.P&I D i
11.5 St Lucie Unit 1, 8770-G-078, Sheet 130, Revision 11 11.6 Combustion Engineering Inc, Minutes of Meeting at Fhnen, July 22,1970 11.7 ASME/ ANSI B16.5-1988, Pipe Flanges and Flanged Fittings 11.8 CWD 8770-B-327, Sheet 289, Revision 12 r
11.9 Ins:ruction Manual 8770-8688, Revision 7 11.10 Emdrac 8770-5628, Revision 0 11.11 Instruction Manual 8770-8941, Revision 17 1
11.12 Stress Calculations a.
SI-676 Part A, dated 01/15/76 b.
SI-676 Part B, dated 10/14/77 c.
SI 2000, dated 06/20/84 11.13 Stress Isometrics a.
SI-199-1, Rev.12 b.
SI-199-4, Rev.10 c.
SI-199-5, Rev.12 d.
SI-199-16, Rev.13 e.
SI-199-17, Rev.13 f.
SI-199-20, Rev.11 g.
BCS-125-1-382, -385, -387, -389, Rev. 7 11.14 Emdrac Drawing 8770-11312, Revision 0 11.15 PC/M 148-185, Otifice Size Calculations for Caustic to Containment Spray System, Revision 4 11.16 FPL Calculation PSL-1FJM-90 050, Revision 0
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i JPN-PSL-SEMP-94-076 Revision 0 Page 1 of 2 l1 l
l Engineered Safeguards Suction Piping l
Minimum Wall Calculation J
I ne largest diameter piping was selected (Reference isometric drawing 877Mi-12S; sheets SI-N-I and CS K-2) for determination of minimum piping wall thickness oflines I-24-SI-504, CS-3, CS-5, CS-41 and other 1/4" wall 24 inch stainless steel pipe when subjected to an internal l
pressure of 250 psig. Per the isometric drawing, the piping design pressure is 60 to 80 psig.
I De code of record is ANSI B31.7 Class 2.
Per USAS B 31.7 (1969 Edition) chapter 2-II, the minimum thickness of a pipe wall required for design pressure shall be in accordance with Division 104 of USAS B31.1:
t, =
PD_
+
a 2(SE + Py) where:
t, = the minimum required wall thickness, in.;
P=
250 psi. This conservatively bounds Containment Spray pump discharge pressure, neglecting piping friction losses due to the spray additive system eductor design flow rate of 128 gpm.
D. = outside diameter of pipe, in.
.= 24.00 in. per Cranes 410, Appendix B for a 24 in. pipe.
SE = maximum allowable stress in material caused by internal prenare.
= 16,600 psi per USAS B31.7, Appendix A, Table A.8, for A-376 TP 304 SS tubular seamless material at 300 *F. Piping was procured seamless per Specification Ebasco 62-72 Appendix A for pipe code SS-1 or SS-2.
a=
an additional thickness to: (1) compensate for material removed during threading, (2) provide for corrosion and/or erosion, and (3) to provide for structural strength of the pipe.during erection, in.
0.0024 in, corrosion allowance per Ebasco Mechanical Engineenng
=
Design Guide, Appendix D, for Safety Injection Piping. No allowance required for pipe threading or construction since piping is welded and already installed.
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y=
a coefficient having values given in USAS B31.1 Table 104.1.2(a)2.
0.4 l
=
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( = (250)(24)
+
0.0024 2((16,600 + (0.4)(250))
t, = 0.182 inches for straight pipe and elbows.
CONCLUSION:
Since the 24 inch piping is nominally 0.250 inches thick, there are no operability concerns for subjecting this piping to 200 psig internal pressure.
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ASNE E16 5 88 M 2595512 0013%21 m l
t ASM E/ ANSI 816. 5-1988 PtPE FLAN 0ES AND FLANGED FITTINGS O
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TABL PRESSURE-TEMPERATURE RA'ilNGS" fressures Are in pw.g /
' D si;460 '
Material Group No.
1.1 1.2 1.3 1.4
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%Mo 1Mo
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- 20 to 100 285 290 265 235 265 290 290 290 290 290 275 275 230 275 275 260 260 200 250 250 250 215 250 250 280 200 260 260 235 240 195 235 245 230 230 300 230 230 230 210 230 230 230 230 230 230' 205 215 175 210 225 220 220 400 200 200 200 200 200 200 200 200 200 200 180 10"-
160 tio 200 200 200 500 170 179 170 170 170 170 170 170 170 170 170 170 145 170 170- 170 170 600 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 650 125 125 125 125 125 125 125 125 125 125 125 125 125 125 125 125 125 700 110 110 110 110 110 110 110 110 110 110 110 110 110 110 110 110 110 750 95 05 95 95 95 95 95 95 95 95 95 85 95 95 95 8b PT 800 80 80 80 80 80 80 80 80 80 80 80 80
,,80, 80 80 80 80 850-65 65 65 65 65 65 65 65 65 65 65 65 65 65 65 65 65 900 50 50 50 50 50 50 50 50 50 50 50 50 50 50 50 50 950 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 1000 20 20 20 20 20 20 20 20 20 20 20 20 20 20 20 10 O
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Datep-4kccaber 6,1994 D. A. Sager hh[ Department:
St. Lucie Plant From:
Vice President
Subject:
PROELDI IHE NQ M459 Ud.1 In accortlance with procedure IP-701, attached is the St. Imcie Plant Problem Repxt on the following subject:
CXNTAINM!NT REEGRIIY OUISIDE & FSAR ASSUlWIKNS UNDER IlMTIH) ORCUM!ifANCES DUE 10 lESKW HRROR Attached is the NP-700 Problem Report generated on this event.
DAS:by cc:
R. A. Block - JPN C. L Burton - PSL R. W. GriiG JE M. Snyder - PSL l
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l TITLE (4) CcIltalDITent intigp.Aty outside of FSAR asstripticria under 122lltec, circumstances rina to design error.
EVENT DATE (5)
LER TRAGER (6)
REPERT CATE (T)
OflER FACILITIES IINOLW (8)
FACILITY NAME DOCKET mseER SEQUENTIAL REVISION MouTN DAY YEAR N/A j
MrvlTH DAY YEAR YEAR NUISER E88ER I
l 10 23 94 94 006 1
12 2
94 '" "^"' N/A l
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OPERAT!bG 1
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MtI)E (9) 20.402(b) f
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50.73(e)(2)(vit)
OTHER Polder LEVEL (10) 20.405(e)(1)(tti) 50.73(a)(2)(1) 50.73(a)(2)(vi t i )( A)
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I LICENSEE CENTACT FCR TNIS LEst (12) j TELEP10tNIE IR8WER (incLLate Aree Cade) i IsAME Michael J. Snyder, Shift 'ntchnical Advisor (407) 465-3550 i
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l (Ltmit to 1400 spaces, i.e., approntantely 15 eingle spaced twitten lines) (16) i A85TRr;T I
on october 20, 1994, Unit 1 was in mode 1 operating at 100% steady state power.
j Differential pressure testing of a trotor ted valve resulted in the lifting of Core Cooling System (ECCS).
a suction supply bandar relief valve for that this relief valve could on 23 october, an engineering evaluation conf-lift under certain accident ccriditions and result in surtp inventory loss in excess of design basis into the Reactor Auxiliary Building, i
i The root cause of the deficiency was design error in the Iodine Removal Sydtem.
l A comtrun header in that system permitted cross train pressurization of an idle J
surizaticr1 of the ECCS suction bandar and the I
Containment Spray purtp, ief valve en that bandar. 'nlis design deficiency had potential to lift the existed since the Iodine Rertoval System was installed in 1978.
i i
- 1) The relief valve path in the Iodine Removal system was Corrective actions:
isolated. 2) FPL Engineering evaluated the effects of increased pressure in the j
'Ihe two reliefs were then ECCS suction header with satisfactory results. 3)
'nie Architect Engineer of the Iodine Removal System was informed of performed. 5)
- 6) A. hit 1 and 2 design review indicated no other similar the design deficiency.
problems in the ECCS.
- 7) 'Ihe ccEmon header was physically separated prior to j
restart from the refueling outage.
Il NRC FORM 366 (5 92)
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TEXT (if more sence is respired, use seHttanet copies of NRC Forst 3464)
TwstrWTirf' TON 01P 'N EVIEM' On 20 October,1994, at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> Unit 1 was at 100% power steady state after successful performance of a.B train Ric;h Pressure operations. I==ad4 ately(EIIS:DQ)'. motor operated valve ~(V3662), differential.
Safety Injecticut Generic Intter.89-10, utility maintenance peh - noted test as recJuired water pooling to a floor drain in the pipe tunnel in the Reactor Auxiliary Butiding The source of the water was found to be frczn a reseated A tra.ul relief (EIIS:NS).
valve SR-07-1A, located en the 1A Emergency Core Cooling (ECES) (EIIS:BP) suction piping.
Health Physica PEE J dataminad that the water was from the Refmlin3 Water Tank.
Later that same shift, utility licensed operators determined that the relief'had lifted during the performa-of the valve differentiki test due to a previously zed pa for crees train pressurization. Se ali;
--r for the test t a @ led a flow frczn the d4 d---
of the 1B Omntairunent Spray pump (IIIS:BE) pung through a ecmEnt:1 header in the revea to the suction the 1A ocarmi -smt A review of plant recceds showed l Iodine Ramoval System (EIIS:BE) (See Figure One).
that during this testing, the maxinum pressure at the A ECES suction piping was 85 S e relief setpoint of SR-07-1A psig.
S e design pressure of the line is 60 psig.- - - -
and IB is 60 psig.
was requested to determine potent 5ial adverse effects Cki 21 October, FPL of ovexpressurizing 1A -
suction piping and to review the operability concern related to the potential to lift SR-07-1A during a postulated design basis accident.
On 23 october, preliminary results from that review pronpted operaticris to isolate ptmp and enter its 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Acticri the nan system from the IB Ccritainment Engineering coupleted a calculation statement at 1255 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.775275e-4 months <br />, At 1915 hours0.0222 days <br />0.532 hours <br />0.00317 weeks <br />7.286575e-4 months <br />, which determined that the conpanents whose design pressure had been exceeded were Se sucticri piping and capable of withstanding censiderably higher pressures.nents, therefore, did not suffer However, it was concluded that a design basis scenario existed which could result cw in lifting the relief valve.
In the event of a tulated Ioss of coolant Accident (wCA) concurrent with a Ioss Of Offsite Power (
P) and the failure of crie Emergency Diesel Generator (EIIS:EK) idle train, and after a Reciru.1 to start, SR-07-1A or 1B could owould release 1td Safeguards equipment containment sunp inventory in excess of the Enginee:'n11s would result in a ccriditical external leakage rate of 2 liters per hour.
outside of the design basis of the Engineered Safeguards systems.
Bis design deficiency had existed since the nan system was backfit to Unit 1 in 1978.
On 26 october, Uhit 1 exited the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Iro when the partially disabled Iodine Removal system was fully restored to service. his was dcz1e after FPL Engineering had performed a Safety Evaluaticr1 which concluded the acceptability of disabling After the the two reliaf valves in the ECCS sucticx1 headers as an interim measure.
l sucticn header reliefs were disabled, a leak test cri each ECCS header confirmed the ace (s W.ility of this mode of operation until the ccxmencement of the Unit 1 refu'eling ou'tage scheduled to begin five days later.
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1 NRC* F0W4 36eA G 92)
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l NRCF0mi3Ma u.s. mucLEAR ReaA.ATORT CWell8SION APPROWED BT ces No. 3150-0106 (5-92)
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St. Lucie Unit 1 maan esen 05000335 94
--006--
1 3 OF 8 4
TEXT (If more space is regaired, use additionet copies of MRC Form 3MA) (17) 4 CADSE OF 'ME EVENT ne root cause of this event was a design deficiency in the Iodine Remenral System.
he Iodine RsErrnral system is a subsystem of the Containment Spray system which is i
used to remove post-accident iodine from the cane =4nmant atrresphere following a IDCA by adding controlled aticunts of sodium hydroxide (nan) to the ccn minmant e
spray water.
21s is acccerplished by mainta2.ning the containment spray solution pH l
within specificaticns to achieve rapid absorption of the radio-iodines and to minimize caustic corrosien of materials and protective coatings within the cent =4nmant.
W e specific design error was that the backfit of the nan system in 1978 did not ocnsider the adverse potential consequences of using a conEnan return l
handar from the dis of the two Ocntainment Spray purrps to the two nan eductors located near suction of each containment spray purrp.
n e discovery of t
this event occurred during the GL 89-10 differential pressure testing of the 1B i
Q:ntainment Spray pump cross tie connection to the IB Hip Pressure Safety Injection purrp.
Other plant test and surveillance m a had isolated the NaOH j
system and therefore had not detected this design d wciency.
fi ANnlMBIS OF EVENT:
Gabl'e'to the NRC under W e postulated lifting of SR-07-1A or 1B is 110CFR50.73.a.2.ii as "Any event or ccnditic plant being in a ccnditicn that was outside the design basis of the plant."
g j
l he purpose of the Ccxitainment Spray system is to prevent the ccntainment vesselle from exceeding its design pressure of,44 peig following a LOCA, assuming a s active or passive failure.
D e Centainment Spray system consists of two radi nt trains. ne heat rerreval capacity of either train is adequate to keep ccntainment pressure and terrperature below design values. W e purpose of the ECCS sucticn relief valves, SR-07-1A and 1B, is to provide relief capability between the low pressure sucticn piping and the higher pressure portion of the sucticn piping used for shutdown cooling.
Iow Pressure Safety Injecticn purrp (LPSI) (EIIS:BP) check valves and ECCS sucticn bandar relief valves were installed to protect against leakage across motor operated isolaticn valves or the failure to isolate this he ECCS portion of the low pressure system prior to initiating shutdown cooling.
suction header relief valves are one and one-half incft reliefs.
l One design basis scenario of ccncern is a large break ICCA with a Containment spray Actuation Signal and a IDOP coincident with one Emergency Diesel Generator failing his would result in cross train pressurizaticn to the non-nmning to operate.
ECCS train, and open a sucticn baadar relief valve. After an RAS, ccntainment surrp inventory release would be in excess of the Engineered Safeguards equipment We external leakage rate of 2 liters per hour assuriptions (FSAR sectica 15.4.1).
maximum leakage rate from the relief would be limited by the design flow of the NaOH spray additive system eductor at 128 gallons per minute.
NRC FORM 366A (3 92)
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w,.s.. e w utC FORM 3MA U.S. NUCLEAR BEalLAftsY Cl3Oll581th AP900WB SY Se NO. 3150 0106
' d (5 92)
EXPIRES 5/31/95 EST!NATED BURDEN PER RESPONSE TO CIBPLY WITN j
TMis INFORMATION COLLECTION RESEST: 50.0 mRS.
j
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nevi m mmen uien St. Incie Unit 1 4 OF 8 05000335 94
--006--
1 I
TEXT (If more space is reeJired additionel copies of NRC Form 3MA) (1; }
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Rta17.Y11tTSI OF 'INEW EVEBar (rwwihsad) ;
i Two additional scenarios which result in relief valve lifting were identified. One scenario postulated a large break IDCA and one ccntatinment sung recirculation valve i
failing to automatically cpen after an RAS.
'Ihis would necessitate shutdown of the affected ECCS train to avoid cavitation of punpa jn that train.
'Ihe second i
scenario tulated a small break IOCA which does not actuate cmtainment spray. A relief ve would lift cm an idle ECCS train during recirculation when a ccntainment spray punp is procedurally aligned to a High Pressure Safety Injecticn punp for NPSH enhancement. 'Ihese small and large break IDCA scenarios described above were evaluated for past significance to plant operaticm and safety.
FPL Engineering lated calculaticms to show that the containmisnt spray system's ability to supply ccmtainment with the design flow rate of at least 2700 gallens per minute was not cenpromised with the diversion through the idle spray pLxlp's bandar and out an ECCS sucticm bandar relief under the design basis scenarios of acrvwen.
An additional calculaticn determnad that the conpanents whose design pressure had been avrmadad during the M W differential pressure testing r.n 20 October were capable of withstar ccmsiderably higher pressures and therefore did not suffer any damage as a re t.of this event.
A review of flooding effects in the ECCE punp rooms.was performed.
Pbr 1 break IDCAs, an RAS will stop LPSI punp(o which provides an ECCS bandar vent path to the Reactor Coolant System (RCS) EIIS:AB), all the sucticm relief to resent and terminate flooding.
For small break IOCAs wi t Ctmtainment Spray actuated, if the size and location of the RCS break allows Shutdown Cooling (SDC) (EIIS:BP) to be placed in service prior to RAS, the idle ECCS train would not be pressurized.
If during a small break IOCA the SDC. system could not be placed in service prior to RAS, then Emergency Operating Pr=d"re 3 (EOP-3), Na of M ane Accia.ne j
directs the operators to align a Containment Spray pung to a High Pressure Safety In]ecticn punp for NPSH purposes. 'Ihis all t would open an idle train's ECCS suction relief. 'Ihe voltane of water from relief valve could be conta2ned for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> without disabling the equipment in the idle ECCS train, and would be contained for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> without adversely affecting the ECCS punps and safety i
related egulpment on the operating ECCS train.
'Ihis is a reascnable time for which procedural and lant staff initiated contingencies would be inplemented to diagnose and mitigate fl effects in an ECCS punp room.
In any IDCA scenario, operators would be erted to flooding by centrol room annunciatim of the ECCS EOP-3 requires operators to check the status of these four sunp level nonitors.
suno monitor alarms; then to investigate and attenpt to isolate the leakage causing high level alarms in either of the two sunps.
Prior to RAS, operators may have enough time to disable the relief by installing a gagging device ucrated on theback to the Reactor Ccntainment Building (RCB) (EIIS:NH)y punping down the ECCS room valve.
Af ter RAS, procedural guidance for resrotel via a dedicated sunp pung system However, this system would be unavailable during a is a specific step in EOP-3.
In the event that these success paths were not inplemented, the ensite IDOP. ency respcnse organizaticn would have sufficient time to diagnose and, mitigate 9thelocxiing in the ECCS punp rocrn prior to rea the opposite ECCS train.
More t upon a high (greater inportantly,)since the ECCS sucticn relief lift isRCB pressure or centinued Containman than 37 psig unlikely that unmitigated flooding in an affected ECCS punp room would continue for more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Nec Foam 3MA (5 92)
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1 NRC F(stM 3eA u.$. NUCLEAR REalLAf0RT Colell8SION APPRtMD BY Om No. 3150 0106 3
3 (5 92).
EXPIRES 5/31/95 ESTIMATED SURDEN PER RESPlataE TO Ct3 PLY WITN
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J THIS INFORMATION COLLECTION RESEST: 50.0 NRs.
FORWARD ColetENTS REGAR0 LNG EmEN FSilRATE TO LI M E W f M r.( E ).
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St. Lucie Unit l
'5 OF 8' 05000335 94
--006--
1 TEXT (le.or..
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This review also concluded that the unaffected train's safety related equi;xnent located throughout the Reactor Auxiliary Building would operate within the bcamda of j
their environmental qu'ilifications in the event of an ECCS suction relief lift.
Leakage from ECCS ccripcnents dur' a IDCA and recirculation phase provide a source to the containment.
All ECCS cw ents of fissicn product leakage ext i
containing recirculating surp water are withiri the controlled ventilaEcn area served by the ECCS area ventilaticn system. This safety related system processes mnts through a charcoal filter before release to the leakage from ECCS c +lant-vent.
Relief valves SR-07-1A and 1B both relieve in i
atmosphere via the p emp=vtments which are served by the ECCS ventilaticn system.
FSAR sectica 15.4.1.7 describes the assumpticms for determining the offsite dose costponent frcm ECCS leakage. 'Ihe offsite and onsite dose consegaences of the relief valve's leakage rate were analyzed using the source term factoring described in NOREG 1465 j
and by increasing the particulate filtration efficiency to operaticnal values.
1 Results from that analysis showed that the offsite dose consequences from the three i
postulated scenarios would not exceed 10 CFR Part 100 Guidelines, and that the onsite dose ccnsequences would not exceed 10 CFR Part 50 General Design Criteria.
'Iherefore, the health and safety of the public were not affected by this condition.
i CORRECTIVE ACTICMS:
- 1) As an interim measure, operaticns isolated one train of the Iodine Removal i,
system to preclude cross ccnnecting the ECCS baadars and lifting the sucticn l
reliefs.
2)
FPL Engineering evaluated the effects of exceeding the design pressure of the ECCS suction bandar during this event and found that the cwpents were capable of withstanding ecnsiderably higher pressures and had not been overpressurized or suffered damage.
3)
FPL Engineering perforned a Safety Evaluation which determined that the ECCS suction headers could withstand ressurization up to the Ccntainment Spray punp j
discharge head ccncurrent with disabling of the relief valves.
4)
As an interim measure, the two ECCS sucticn baader relief valves were disabled.
5)
'Ihe Technical Staff performed satisfactory leak testing of the ECCS suction l
l header.
6)
Operations fully restored the Iodine Removal system to service within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I.CO time limit.
t 7)
A review of the Unit 2 Iodine Removal system verified that it was ncit susceptible to a similar design deficiency as noted cn Unit 1.
I NaC FORM 366A (5 H )
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NunsER wuseER St. Lucie Unit 1 6 OF 8 05000335 94
--006--
1 rExt cit mor.
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, neswr na Acricnasi (eemH ernad) :
1 8)
A design review was conducted on the Unit 1 and Unit 2 ECCS piping used during l
injection and recirculation modes of operation. No other paths for cross flow were 1
identified to be outside the license design basis.
1 l
9)
A perTnanent modification to physically separate the conman NaOH eductor was j
accomplished prior to unit restart from the refueling outage. (See Figure Two) 10)
FPL Engineering has informed the plant Architect Engineer of the this design j
4 deficiency, 11)
Iessons learned from this event were shared on the INPO's Nuclear Network.
i l
@ITIGULL LNGRDLTICM System Tdantifirmeion:
Iodine Removal System: NaOH System with eductors Architect Engineer: Raytheon (Ebasco)
I I
i NSSS: Ccebustion Engineering l
Pravious similar Events:
A previous LER at St. Lucie related to design deficiencies which resulted in a condition of the plant being outside of the design assumptions in the FSAR is LER 335-80-27, "Unanalyzed Boron Diluticn Transient."
l l
NRC FORM 366A (5-92)
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DEC-e7-95 THU 11840 IMACE NGNT CORP 4074669991 P.13 9
Inter 0ffice Correspondence FPL JPN-SPSL-95-0414 Tot S. A. Valdes Dates October 10, 1995 St. Lucie Plant 4% V> H % >
From D. J. Denver Department: ENG/PSL Nuclear Engineering subjects ST. LUCIs PI. ANT UNIT 1
,,,JPN= Psi,.SRIP-95425, 5tsT. 1 PLANT OPERATION AT 2225 PSIA NOMINAL OPERATING PRESSURE PROJECT # N/A PILE RE
Reference:
Siemens Letter RIW:95:030, Wescott to j
Knuckles, dated March 9, 1995.
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Engineering evaluation JPN-PSL-SENP-95-025, Rev.
1 has been issued for uwe.
This evaluation provides for operation of Unit 1 at the existing technical specification 3.2.5 pressurizer pressure limit of 2225 psia.
Although within existing accident analyses, operation at this lower pressure slightly reduces the margin available to DNB.
Per the referenced analysis, a 25 psi drop in pressure equates i
to a power penalty of -1.7%
(for the most limiting case).
Therefore, operation at a reactor power level of 98.3% and 2225 psia provides for the same margin to DNB as operation at 100%
and 2250 psia.
However, sufficient DNB margin is available to operate at 100% and 2225 psia as demonstrated in the attached evaluation.
ACTIONS None 1
If you have any questions please contact Chris Wasik at x7491, i
i DJD/CJW 1
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