ML17227A389
| ML17227A389 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/21/1992 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17227A388 | List: |
| References | |
| NUDOCS 9204230201 | |
| Download: ML17227A389 (143) | |
Text
St. Lucie Unit
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2 Docket No. 50-389
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Proposed License Amendment Safet In'ection Tank Minimum Pressure Reduction ATTACHMENT 1 St. Lucie Unit 2 Marked-up Technical Specification Page 3/4 5-1
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3t'0';5-" 'tNGBCY'OttE~CNRING SYSTEMS ECCS 3/4. 5. 1 SAFETY INJECTION TANKS iV. ~ s LIMITING CONDITION FOR OPERATION
~ I 3.5; 1 Each Reactor Coolant System safety injection tank shall be OPERABLE with:
a.: The, isolation valve..open,
- b. A contained borated water volume of between 1420 and 1556 cubic feet, I
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- d. A niItrogcn~cover.-pressure of between @'nd '650 psig.
APPI ICABILITY: MOOES 1; 2, 3", and 4". Sob.
ACTION:
With one safety injection tank inoperable, except as a result of a closed isolation valve, restore the insuperable.tank to Awithin 1"ho'uh,or. be 'in .af;ljast HdT sTACIBY 'within the next; QPERA8LE.')atua 6'o'urs and in HOT SHUTDOWN"wi4hiri the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
+th 'one,'sat'ety ing'ection tang, inoperabIe'.due,.'to)gei~1atlqn valve betng chas'ed, either, .iaeedfgCqty,open,'he "isolatian.vie er .Pe,.in at least'HOT STANDB'(within 1'our an4 tje in'OT,SHUTGGMtfmithin the next. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />...
~ l SURVEILLANCE. RE UIREMENTS.:
4.5;,l-..lI Each safari@;injection tank shall be demonstrated.OPFRABI.E; ...
- a. At least once per 12 hours by:
- l. Verifying (by the absence of alarms) the contained borated water volume and. nitrogen cover pressure in the tanks, and 2;" Verifying'hat 'each safety injection tank isolation valve is open.
th pressur Zer press'ut"e,'..cfihaterI thai or equal to 1750 psia.'hen pressur-izer pressure is 1'ess than 1750 psia",at least three safety injection. tanks shall be OPERABLE, each with a minimum pressure of., 235 psig and a maximum pressure of 650 psig and a contained'ater vol'ume of between 1250 and 1556 cubic feet with a.boron concentration of between,1720 and 2100 ppm of boron.
With all four s'afety injection tariks 'OPERAHL'E, each tank shall have a minimum pressure of 235 psig and a. maximum pressure of 650. psig ante a caqtaiqed .water volume of Between 833'nd.'1556 cubic feet with a boih concentratian of between 1720 and 2100 ppm of b'oron. In MODE 4 with pressuriziir pressUre less than 276 psia, the safe.'y injection tanks may be isolated.
ST. LUCIE " UNIT 2 3/4 5"1 Amendment No.40 I
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St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Safet In ection Tank Minimum Pressure Reduction ATTACHMENT 2 SAFETY"ANALYSIS Introduction The proposed change to the St. Lucie Unit 2 ,Technical Specifications reduces the safety injection tank (SIT) minimum pressure from 570 psig to 500 psig. The proposed change modifies Technical Specification 3.5.1.d, Safety Injection Tanks, for St.
Lucie Unit 2.
The proposed change provides the benefit of a greater differential pressure margin between the SIT operating pressure and the SIT relief valve pressure setpoint (669 psig). Reducing the SIT Limiting Condition for Operation (LCO) to 500 psig means the minimum SIT pressure would be approximately 75 percent of the re'lief valve pressure setpoint which represents a 10 percent increase in margin when compared to current conditions. This additional margin lessens the potential for SIT relief valve leakage that has impacted plant availability in the past.
Discussion The SITs ensure that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the reactor coolant system (RCS) pressure falls below the pressure of the SITs. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on SIT volume, boron concentration, and pressure ensure the assumptions used for SIT injection in the safety analysis are met.
By comparison, St. Lucie Unit 1 SIT nitrogen cover pressure is normally set between 220 psig and 225 psig, and the Unit 1 Technical Specifications require the SIT pressure to be between 200 psig and 250 psig. St. Lucie Unit 1 has 14 x 14 fuel assembly array while St. Lucie Unit 2 has a 16 x 16 fuel. The higher pressure of the St. Lucie Unit 2 safety injection tanks was employed to gain additional LOCA margin above the margin gained by the fuel assembly array change.
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Section 6.3.3.1 of the St. Lucie Updated Final Safety Analysis
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Report (UFSAR) lists the following Emergency Core Cooling System (ECCS) design criteria from 10 CFR 50.46:
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(1) The calculated maximum-fuel element temperature shall not exceed 2200'F.
(2) The calculated total oxidation of the cladding shall nowhere exceed 17 percent of the total cladding thickness before oxidation.
(3) The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 1 percent of the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(4) Calculated changes in core geometry shall be such that the. core remains amenable to cooling.
(5) After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. is the Technical Evaluation, Reduction of Safety Inj ection Tank Pressure Minimum Setpoint for St. Luci e Unit 2 Nuclear Power P2ant, prepared by ABB Combustion Engineering.
provides the justification for a decrease in the St. Lucie Unit 2 It minimum SIT pressure from 570 psig .to 500 psig. Technical justification for operation of St. Lucie"Cycle"7"'at a-reduced SIT minimum operating pressure of 500 psig is provided by ECCS performance evaluations of the small break loss-of-coolant accident (SBLOCA). Since the large break loss-of-coolant accident (LBLOCA) analysis of record was performed with,an SIT pressure of 200 psig, the proposed minimum SIT pressure of 500 psig is bounded. These performance evaluations demonstrate acceptable conformance with 10 CFR .50. 46.
A review of the non-LOCA design bases events, in'ttachment 4, shows that =none of these analyses calculate or-credit SIT injection into the RCS. Therefore, the reduction of SIT-minimum operating pressure setpoint has no impact on postulated non-LOCA design bases events.
Although the station blackout event (SBO) is outside the design bases for Unit 2, an analysis was performed as a condition of license at the request of the NRC. This analysis appears as Section 15.10 of the UFSAR. Attachment 4 provides' technical
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evaluation of the SBO event at the proposed minimum operating pressure of 500 psig. This .evaluation demonstrates that the original conclusions presented in Section 15.10.5 of the UFSAR have not been adversely affected.
The slight increase in the differential pressure between the RCS and SITs may have two opposing effects on backleakage from the RCS into the SITs. First, the increased differential pressure will tend to seat the check valves more firmly, and thus may cause a slight decrease in the probability of backleakage into the SITs.
Second, the slight increase in differential pressure between the RCS and the SITs may slightly increase the rate of backleakage into the SITs, should backleakage occur. Such back-leakage would affect the required SIT boron concentration and level. There is no history of back-leakage into the SITs at St. Lucie Unit 2 or Unit
- 1. As a change in level would alert the operators to any accompanying reduction in boron concentration in the SITs, there is essentially no probability of such back leakage contributing to an event. Therefore, the reduction of SIT minimum pressure will have no significant effect on the probability or consequences of back-leakage from the RCS into the SITs.
CONCLUSION The acceptability of the proposed reduction of the SIT minimum operating pressure depends on the impact on the postulated SBLOCA, the postulated LBLOCA, and the postulated SBO.
The SBLOCA limiting break size and associated peak cladding temperature strongly depend on the SIT pressure. Reducing the SIT minimum operating pressure to 500 psig for St. Lucie Unit 2 increases the SBLOCA limiting break size from 0.0375 ft2 to 0.0450 ft and increases the peak cladding temperature from 1771'F to 1905'F. This analytical evaluation for SBLOCA was performed using ABB Combustion Engineering's NRC approved proprietary small break evaluation model, Calculative Methods for the CE Small Break LOCA Evaluation Model, CENPD-137(P), dated August 1974 and CENPD-137(P)
Supplement 1, dated January 1977. The analysis was performed consistent with the requirements of 10 CFR 50 Appendix K. This analysis demonstrated acceptable conformance with 10 CFR 50.46.
The LBLOCA analysis of record is for cycle 3 which bounds later cycles. The LBLOCA analysis was redone for cycle 2 when the core power was increased from 2560 MWt to 2700 MWt. -The LBLOCA analysis was redone again for cycle 3 with revised steam generator tube plugging limits. This LBLOCA analysis was performed with a SIT pressure of 200 psig, a condition which conservatively covers the 500 psig SIT pressure used in the SBLOCA analysis. The peak cladding temperature for the LBLOCA analysis is 2107'F. The SBLOCA analysis with reduced SIT minimum pressure remains less than the current LBLOCA peak cladding temperature of record. Therefore, the e
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LBLOCA calculation remains the. limiting analysis of..record for ECCS performance evaluation. This LBLOCA analysis also shows acceptable
.conformance to the 10 CFR 50.46 acceptance criteria for ECCS performance in support of the reduced value of SIT minimum pressure of 500 psig.
Attachment 4 includes a review of SBO, which. is outside the design basis for St. Lucie Unit 2 but is part of the licensing basis. The SBO event postulates a four hour time period for restoring AC power during which primary subcooling and natural circulation must be maintained. The SBO analysis in UFSAR Section 15.10 credits some SIT injection late in the transient. An assessment of this analysis,.concluded that the proposed reduction in the minimum SIT pressure would delay the SIT discharge. However, sufficient inventory remains in the RCS through the SBO to prevent voiding in the RCS loop and thus to prevent a loss of natural circulation.
The conclusions presented in the UFSAR regarding maintaining natural circulation and core subcriticality are unchanged.
Therefore, the reduction of the SIT minimum pressure setpoint to
.-500'sig has no impact on the SBO event.
Non-LOCA design bases events have been reviewed to evaluate the impact of decreasing the SIT minimum pressure to 500 psig in Attachment 4. For St. Lucie Unit 2, none of the non-LOCA events calculate or credit SIT injection into the RCS. Therefore, the reduction in SIT minimum operating pressure setpoint has no impact, on other postulated non-LOCA safety analyses.
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St. Lucie Unit 2
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Docket No. 50-389
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Proposed License Amendment Safet In'ection Tank Minimum Pressure Reduction ATTACHMENT 3 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that a request for amendment involves a no significant hazards consideration are included in the Commission s regulation, 10 CFR 50.92, which states that no significant hazards considerations are involved operation of the facility if in accordance with the proposed amendment the would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Each standard is discussed as follows:
(1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.
Reducing the safety injection tank (SIT) minimum pressure does not involve a significant increase in the probability of a
-.loss-of-coolant accident -(LOCA),- since -the -SITs --are--passive systems and have no effect on the reactor coolant system (RCS) until after the depressurization of the RCS due to a LOCA.
Reducing the SIT minimum pressure to 500 psig for St. Lucie Unit 2 increases the small break loss-of-coolant (SBLOCA) limiting break size from 0.0375 ft to 0.0450 ft and increases peak cladding temperature from 1771'F to 1905'F for the new limiting break size. The calculated peak cladding temperature (1905'F) for the SBLOCA analysis with reduced SIT minimum pressure remains less than the current large'.break loss-of-
.coolant (LBLOCA),analysis. peak cladding .temperature. of.2107'F..
This LBLOCA analysis was performed assuming a SIT pressure of 200 psig, a condition which conservatively covers the proposed minimum SIT pressure of 500 psig. Therefore, although the consequences of a SBLOCA are increased slightly, the LBLOCA calculation remains the limiting analysis of record for emergency core cooling system (ECCS) performance evaluation.
The LBLOCA analysis of record is for cycle 3 which has been shown to bound later cycles. This LBLOCA analysis also shows acceptable conformance to 10 CFR 50.46, Acceptance Criteria for ECCS performance for light water nuclear power reactors, in support of the proposed minimum SIT pressure of 500 psig.
This ECCS performance evaluation for St. Lucie Unit 2 was performed consistent with NRC approved methodology and 10 CFR 50 Appendix K criteria. The LBLOCA analysis was performed assuming a SIT pressure of 200 psig, therefore, there is no increase in the consequences of a LBLOCA due to reducing the minimum SIT pressure to 500 psig.
Non-LOCA design basis events have been reviewed to evaluate the impact of decreasing the minimum SIT pressure to 500 psig.
For St. Lucie Unit 2, none of the non-LOCA events calculate or credit SIT injection into the RCS, since none of the non-LOCA accidents result in RCS depressurization below the SIT maximum pressure setpoint. Therefore, the reduction of the SIT minimum pressure setpoint has no increase in the consequences
. of non-LOCA design bases events due to reducing the minimum SIT pressure to 500 psig.
The station blackout event, (SBO), which is presented in Section 15.10.5 of the UFSAR, has been reviewed to evaluate the impact of decreasing the SIT minimum pressure to 500 psig.
With the SIT pressure reduced to 500 psig, SIT injection is initiated prior to the occurrence of voiding in the RCS loops, thus preventing a loss of natural circulation.
Therefore, reducing the St. Lucie Unit 2 Technical Specification limit for SIT minimum nitrogen cover pressure from 570 psig to 500 psig does not involve a significant increase in the probability. or *.consequences=='of-.an=-.accident previously evaluated.
(2) Use of the modified specification would not create the possibility of a new or different kind of accident from any previously evaluated.
There are no additional failure modes for the SITs due to reducing the nitrogen cover pressure from 570 psig to .500
.psig. The SITs"are-passive"systems and have-no.effect-on -the
,RCS until after the depressurxzation of-.the-RCS.,due to.a .LOCA.
Therefore, reducing the St. Lucie Unit 2 Technical Specification limit for SIT minimum nitrogen cover pressure from 570 psig to 500 psig does not create the possibility of a new or different kind of accident from any previously evaluated.
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(3) Use of the modified specification would not involve a significant reduction in a margin of safety.
The SBLOCA analysis with the SIT pressure of 500 psig satisfies the criteria of 10 CFR 50.46 and remains bounded by the LBLOCA analysis of record. The LBLOCA analysis was
.performed assuming a.SIT pressure of 200 psig. Review of the SBO analysis with the SIT pressure of 500 psig demonstrates that the original conclusions, presented in UFSAR Section 15.10.5, have not been adversely affected.
Therefore, reducing the St. Lucie Unit 2 Technical Specification limit for SIT minimum nitrogen cover pressure from 560 psig to 500 psig does not involve a significant reduction in a margin of safety.
Based on the above, we have determined that the proposed amendment does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the probability of a new -or different kind of accident from any previously evaluated, or (3) involve a significant reduction in a margin of safety; and therefore does not involve a significant hazards consideration.
+I St. Lucie Unit
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, Proposed License Amendment Safet In'ection Tank Minimum Pressure Reduction ATTACHMENT 4 TECHNICAL EVALUATION REDUCTION OF SAFETY INJECTION TANK PRESSURE MINIMUM SETPOINT FOR ST. LUCIE UNIT 2
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OPS-92-0385 ENCLOSURE 1 TECHNjtCALJUS'IXFICATION for REDUCTION OF SAF ETY INJECTION TMW PRESSURE SETPOINT TO 500 PSIG for ST; LUCIE VI.'GT 2 NUCLEAR POWER PLANT Prepared for Florida Power 4 Light
'y Operations Analysis ABB Combustion Engineering Nuclear Services MARCH 1992 Combustion Engineering, Inc.
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This report provides technical justification for a decrease in the St.
Lucie Unit 2 minimum operating safety injection tank (SIT) pressure from 570 psig to 500 psig. Reducing this Technical Specification Limiting Condition for Operation (LCO) value provides a greater differential pressure margin between the SIT operating pressure and the SIT relief valve pressure setpoint (669 psig). This additional margin lessens. the potential for challenges to the SIT relief valve and concurrent relief valve leakages that have impacted St. Lucie Unit 2 availability'n the past. Technical justification for operation of St. Lucie Unit 2 Cycle 6 at the reduced SIT minimum pressure of 500 psig is provided by ECCS performance evaluations of the small break loss-of-coolant accident.
Reducing the SIT minimum operating pressure to 500 psig increases the SBLOCA limiting break size from 0.0375 ft to 0.045 ft and increases the peak cladding temperature from 1771'F to 1905'F. This increased PCT for SBLOCA remains less than the limiting ECCS performance analysis large break LOCA PCT of 2107'F. The LBLOCA analysis was performed with a minimum SIT pressure of 200 psig, therefore, the LBLOCA calculated results already conservatively cover operation at a minimum SIT pressure of 500 psig. These ECCS performance evaluations demonstrate acceptable conformance with 10CFR50.46. A review of the non-LOCA design basis events shows that none of these safety analyses credit SIT injection into the RCS to meet the acceptance criteria. Therefore, the reduction of the SIT minimum opeFating pressure setpoint has no adverse impact on the non-LOCA design basis events. An assessment of the Station Blackout Event shows that even with delayed SIT actuation due to the reduced minimum operating pressure setpoint, primary reactor coolant subcooling in the hot legs and reactor coolant system natural circulation are maintained throughout the assumed four hour time period for restoring AC power during this postulated event.
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- 2. 1 Introduction and Summary 2.2 Method of Analysis 2.3 Results 2.4 Conclusions 5Y@7%NNCCVN9KNQMxYNN9%NNNC6 48 51 4.1 Non-LOCA Design Basis Events 51 4.2 Station Blackout Event 55 60
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This report provides technical justification for a decrease in the St.
Lucie Unit 2 minimum operating safety injection tank (SIT) pressure from 570 psig to 500 psig. Reducing this Technical Specification Limiting Condition for Operation (LCO) value provides a greater differential pressure margin between the SIT operating pressure and the SIT relief valve pressure setpoint (669'sig). Reducing the LCO to 500 psig means
'hat the minimum SIT pressure for operation would be roughly 75% of the relief valve pressure setpoint which represents a 1(N increase in margin compared to current conditions. This additional margin lessens the potential for challenges to the SIT relief valve and concurrent relief valve leakage that have impacted St. Lucie Unit 2 availability in the past.
This report is limited to the engineering evaluation necessary to justify the plant change to reduce the SIT minimum operating pressure setpoint from 570 psig to 500 psig. This report does not include (1) an evaluation of the related instrument (loop) inaccuracies or uncertainties, (2) PC/H, technical specification, FSAR or related change packages, or (3) addressing changes in the SIT maximum operating pressure setpoint.
Primarily, this report addresses the impact of reducing the technical specification,LCO on minimum SIT pressure for the postulated small break loss-of-coolant accident (SBLOCA). The SBLOCA limiting break size and associated peak cladding temperature (PCT) strongly depend on the SIT pressure. Reducing the SIT minimum operating pressure to 500 psig for St.
Lucie Unit f increases the SBLOCA limiting break size from 0.0375 ft to 0.045 ft and increases the PCT from 1771'F to 1905'F. This analytical evaluation for SBLOCA was performed using ABB Combustion Engineering's NRC approved small break evaluation model described in Reference 1. This analysis demonstrates acceptable conformance with 10CFR50.46 which presents the Acceptance Criteria for Emergency Core Cooling Systems (ECCS) for Light Water Nuclear Power Reactors (Reference 2).
The calculated PCT (1905'F) for the SBLOCA analysis with reduced SIT minimum operating pressure remains less than the current large break loss-
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of-coolant accident (LBLOCA) PCT of record. The PCT for the LBLOCA analysis is 2107'F. This LBLOCA analysis was performed with a SIT operating pressure of 200 psig, a condition which conservatively covers the 500 psig SIT pressure used in the SBLOCA analysis. Therefore, the LBLOCA calculation remains the limiting analysis of record for ECCS performance evaluation. The LBLOCA analysis of record is for Cycle 3 which has been shown to bound later cycles through Cycle 6. This LBLOCA analysis also shows acceptable conformance to 10CFR50.46 Acceptance Criteria for ECCS performance in support of the reduced value of SIT minimum operating pressure of 500 psig. This LBLOCA ECCS performance evaluation for St. Lucie Unit 2 Cycle 6 was performed at a power level of 2754 NMt (2700 HMt plus 2X uncertainty) and at a peak linear heat generation rate (PLHGR) of 13 kw/ft.
This report includes a review of the non-LOCA design basis events to I
evaluate the potential impact of decreasing the LCO for SIT minimum pressure to 500 psig. For St. Lucie Unit 2, none of the non-LOCA events credit SIT injection into the reactor coolant system. Therefore, the reduction of the SIT minimum operating pressure setpoint has no adverse impact on the non-LOCA design basis event analyses.
This report includes an assessment of the validity of the station blackout event documented in Section 15.10 of the FSAR (Reference 9) with regard to the impact of reducing the minimum SIT pressure setpoint. In this assessment, the results presented in Reference 9 are used to demonstrate that even with delayed SIT discharge into the RCS resulting from lowering the SIT setpoint to 500 psig, sufficient liquid inventory remains in the RCS to prevent voiding in the RCS loop and loss of natural circulation.
Therefore, this assessment shows that the conclusions presented in Reference 9 regarding maintaining primary natural circulation and core subcriticality for the four hour duration of the postulated event are unchanged with the reduced SIT pressure setpoint of 500 psig.
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2.1 Introduction and Summar This section presents the results of the ECCS performance evaluation for the Small Break Loss-of-Coolant Accident (SBLOCA) for St. Lucie Unit 2, Cycle 6, with the minimum operating safety injection tank (SIT) pressure decreased from 570 psig to 500 psig. This ECCS performance evaluation demonstrates conformance with 10CFR50.46 which presents the Acceptance Criteria for Emergency Core Cooling Systems (ECCS) for Light Water Nuclear Power Reactors (Reference 2). The evaluation with 500 psig minimum operating SIT pressure demonstrates acceptable SBLOCA ECCS performance for St. Lucie Unit 2. Sections 2.2, 2.3, and 2.4 present the method of analysis, results, and conclusions, respectively.
2.2 Hethod of Anal sis The calculations reported in this section were performed using ABB Combustion Engineering's NRC approved small break evaluation model described in Reference 1. This method of analysis is the same as that used for the St. Lucie Unit 2 Cycle 6 SBLOCA ECCS performance analysis (Reference 3).
A complete evaluation of SBLOCA involves the use of four computer codes.
Blowdown hydraulics are calculated using the CEFLASH-4AS computer code (Reference ~ Reflood hydraulics are calculated using the COHPERC-II code (Reference 5). Fuel rod temperatures and cladding oxidation percentages are calculated using the STRIKIN-II (Reference 6) and PARCH (Reference 7) codes. Details of the interfacing of these codes are discussed in Reference 1.
For the ECCS performance evaluation with the reduced SIT pressure, only the CEFLASH-4AS and PARCH codes were needed for the analysis. Lowering the SIT pressure while keeping all other parameters and methods the same results in a change in the limiting break size and a change in the
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calculated PCT. The CEFLASH-4AS code is used to provide primary system thermal-hydraulics for a spectrum of break sizes to identify the new limiting break size and the PARCH code provides the hot rod heat-up calculations during the core boil-off period of the transient.
For licensing calculations, the limiting break size is typically the largest break for which the cladding temperature heat-up during core boil-off uncovery is reversed with only HPSI delivery to the primary system and without SIT discharge. In this analysis, PCT occurs just after a brief momentary discharge from the SITs which is followed by HPSI driven repressurization and core recovery. For SBLOCA, the COHPERC-II code is used to analyze the core hydraulic response (recovery) following continuous SIT discharge. The Evaluation Model methodology in Reference I for analyzing core reflood following SIT discharge using COHPERC-II assumes that continuous SIT injection causes all steam in the core two-phase fluid to condense. This condition does not apply for the break sizes of this study where SIT discharge is brief and not continuous. for the break sizes in this SBLOCA study, core reflood from HPSI delivery and brief discharge from the SITs is analyzed adequately with the CEFLASH-4AS code which gives both the thermal and hydraulic response of the core reflood period. Therefore, analysis of core reflood using the COHPERC-II code is not needed for the limited range of break sizes in this study.
In the ABB C-E methodology for SBLOCA, the STRIKIN-II code is used to analyze the hot rod thermal response during the period of forced convection when departure from nucleate boiling may occur. For the break sizes of this SBLOCA study, a cladding temperature increase during the forced convection period due to departure from nucleate boiling does not
,usually occur, as indicated by the 0.0375 ft break results from the analysis performed for Cycle 6 in Reference 3. Since the PCT calculated with the PARCH code occurs during the boil-off uncovery period of the transient, analysis of the early blowdown hot rod response using the STRIKIN-II code is not needed for the limited range of break sizes in this ECCS evaluation. Also, the early blowdown hot rod response has a negligible influence on the boil-off uncovery period when departure from nucleate boiling does not occur.
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1 "The ECCS analysis assumptions for this analysis with reduced SIT pressure are unchanged from the evaluation model assumptions (Reference 1). The worst single failure for analysis of SBLOCA is the failure of one of the emergency diesel generators to start. This failure results in the minimum safety inje'ction available to cool the core. Therefore, based on this assumption, the following injection pumps were credited in the SBLOCA analysis:
'a ~ one High Pressure Safety Injection (HPSI) pump
- b. one Low Pressure Safety Injection (LPSI) pump C. one Charging pump
- d. four Safety Injection Tanks For breaks in the pump discharge leg, it is also assumed that all safety injection flow delivered to the broken cold leg spills out the break.
This results in the following r educed injection flow delivered to the core:
'a ~ 75% of the flow from one HPSI pump
- b. 50X of the flow from one LPSI pump C. 4Ã of the flow from one Charging pump (based on worst flow split)
- d. 10Ã of the flow from three Safety Injection Tanks As described in Reference 1, the SBLOCA analyses assume that offsite power is lost upon reactor trip. As a result, the safety injection pumps were assumed to start after a 30 second delay (for diesel generator startup and load sequenqjgg) following a safety injection actuation signal.
The ECCS performance analysis considered a spectrum of cold leg breaks in the reactor coolant pump discharge leg. As demonstrated in previous analyses, discharge leg breaks are more limiting than other break locations. The spectrum of breaks were selected to show the new limiting break size resulting from the reduction in SIT pressure.
I'tJ The previous limiting break size, 0.0375 ft, is not influenced by the reduced SIT pressure, since the cladding temperature heat-up is reversed by HPSI delivery alone. Larger break sizes depend on SIT discharge to reverse the fuel cladding heat-up. By reducing SIT pressure, these larger break sizes'ust wait for the reactor coolant system pressure to decrease to the lower SIT setpoint pressure before receiving SIT discharge to the primary system. These larger break sizes depressurize more rapidly, lose inventory out the break at a greater rate, and experience greater rates of coolant flashing. But, these larger break sizes also receive more HPSI delivery to the primary system with the lower primary system pressure prior to SIT discharge. While these larger break size cases wait for. SIT actuation at the lower pressure setpoint, the cladding continues to heat-up to higher temperatures.
Eventually HPSI delivery at the lower primary system pressure will be greater than the inventory loss by flashing, core boil-off, and spillage out the break and will reverse the cladding heat-up just prior to or near the moment of SIT discharge, thus establishing a new limiting break size.
Brief SIT discharge to the RCS can occur just prior to reaching the time of peak cladding temperature and may through its impact on the core two-.
phase mixture level terminate the cladding temperature rise and/or may cause the core axial elevation of peak temperature to change to the next highest node. The new limiting break size identified by this analysis is the 0.045 ft cold leg break.
2.3 Results The analysis demonstrated the 0.045 ft break to be the limiting small break with a peak cladding temperature of 1905'F and a maximum cladding oxidation percentage of less than 7%. The results are summarized in Table 2.3-1. The times at which significant events in the performance of the ECCS occurred for each break size are listed in Table 2.3-2. Table 2.3-3 provides a list of the significant parameters and initial conditions used
,in the analysis.
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The transient values of parameters which most directly affect fuel rod performance are shown in Figures 2.3-1 through 2.3-4 (see Tables 2.3-4 and 2.3-5). The following parameters are graphically presented for each break size:
(a) Normalized Total Core Power (b) Inner Vessel Pressure (c) Break Flow Rate Inner Vessel Inlet Flow Rate
'd)
(e) Inner Vessel Two-Phase Mixture Level (f) Hot Spot Heat Transfer Coefficient (g) Coolant Temperature at Hot Spot (h) Hot Spot Cladding Surface Temperature Figure 2.3-5 summarizes the peak cladding temperature results of the spectrum analysis.
The 0.045 ft break was determined to be the limiting small break. For breaks smaller than 0.045 ft core uncovery begins later when the fission product decay heat generation is less and, hence, the depth of uncovery will be less. for breaks greater than 0.045 fthm the resulting system depressurization rate is faster such that the cladding temperature rise is terminated early in the transient by Safety Injection Tanks actuation.
2.4 Conclusions Based on the results of an analysis of a spectrum of small breaks in the cold leg at the reactor pump discharge with the minimum operating pressure of the safety injection tanks reduced from 570 psig to 500 psig, it is
'concluded that operation of St. Lucie Unit 2 Cycle 6 is acceptable. The results of the limiting 0.045 ft small break resulted in a peak cladding temperature of 1905'F and a maximum cladding oxidation percentage of less than 7%, demonstrating the SBLOCA ECCS performance to be less limiting than that for the large break LOCA performance results given in Section 3.0.
A vs
Table 2.3-1 St. Lucie Unit 2 Fuel Rod Performance Summary Small Break LOCA Spectrum with Safety Injection Tank Pressure of 500 psig Peak Clad Local Hot Rod Break Size Temperature Clad Oxidation Clad (ft ) location"'eak axial axial
" Oxidation"'1o)
('F) P) location 0.0375 ft /PD 1771 0.90 5.24 0.90 <0.65 0.0400 ft /PD 1836 0.95 6.73 0.90 <0.81 0.0450 ft /PD( 1905 0.95 6.69 0.90 <0.77 0.0500 ft /PD 1841 0.85 4.61 0.90 <0.53 (a) Hot rod oxidation values are given as a conservative indication of core-wide oxidation (b) PD at Pump Discharge (c) Axial location given as a fraction of active core height 10
~ ~
Table 2.3-2 St. Lucie Unit 2 Times of Interest for Small Break LOCA Spectrum with Safety Injection Tank Pressure of 500 psig (seconds after break)
Hot Spot Break Size LPSI SI~ ~
Peak Clad (ft )
HPSI'"'ump On Pump On Tanks On Temp. Occurs 0.0375 ft /PD 130 (a) 2807 2261.2 0.0400 ft /PD 120 (a) 2408 2351.7 0.0450 ft /PD 110 (a) 2032 2034.3 0.0500 ft /PD 100 (a) 1763 1776.2 (a) Calculation terminated before time of LPSI pump activation.
(b) "Pump On" and "Tanks On" are meant to indicate "RCS Injection Begins"
Table 2.3-3 St. Lucie Unit 2 Small Break ECCS Performance Analysis Significant Parameters and Initial Conditions arameter Values Core Power Level 2754 at 102K of Nominal (MWt)
Core Average Linear Heat Rate 4.90 at 102K of Nominal (kw/ft)
Peak Linear Heat Generation Rate (PLHGR) 15.0 Hot Assembly, Hot Channel (kw/ft)
Core Inlet Temperature ('F) 552.0 Core Outlet Temperature ('F) 603.8 System Flow Rate (ibm/hr) 136.1xlO (1)
Core Flow Rate (ibm/hr) 131.1xlO Number of Tubes Plugged Per Steam Generator 1250
~ Safety Injection Tank (SIT) Gas Pressure (psig) 500 (2) moderator Temperature Coefficient (dp/'F) +0.2xlO Axial Shape Index (ASIU) -0.15 Low'ressurizer Pressure Trip Setpoint (psia) 1650.0 I
Safety Injection Actuation Signal Setpoint (psia) 1500.0 High Pressure Safety Injection Pump Shutoff Head (psia) 1214 (1) System Flowrate consistent with 363,000 gpm (2) LBLOCA uses SIT Gas Pressure of 200 psig 12
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Table 2.3-4 St. Lucie Unit 2 Small Break LOCA Spectrum I'e Si e and Location Abbreviation ~Fi ures 0.0375 ft Break in Pump 0.0375 ft /PD 2.3-1 Discharge Leg 0.0400 ft Break in Pump 0.0400 ft /PD 2.3-2 Discharge Leg 0.0450 ft. Break in Pump 0.0450 ft /PD 2.3-3 Discharge Leg 0.0500 ft Break in Pump 0.0500 ft /PD 2.3-4 Discharge Leg 13
Table 2.3-5 St. Lucie Unit 2 Variables Plotted as a Function of Time for Each Small Break LOCA in the Spectrum Figure Variable Normalized Total Core Power Inner Vessel Pressure Break Flow Rate Inner Vessel Inlet Fl'ow Rate Inner Vessel Two-Phase Hixture Level Heat Transfer Coefficient at Hot Spot Coolant Temperature at Hot Spot Hot Spot Cladding Surface Temperature I
- Refer to Figures 2.3-1A through 2.3-4H.
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The large break loss-of-coolant accident (LBLOCA) analysis of record for St. Lucie Unit 2 is for Cycle 3 (Reference 8). The Cycle 3 analysis has bounded later cycles, namely, Cycles 4, 5, and 6, because all current cycle-specific input data has been shown to be the same or less severe than the Cycle 3 data.
The Cycle 3 input data and results of Reference 8, which apply conservatively to Cycle 6 are presented in Tables 3.0-1 and 3.0-2, respectively. The results of this ECCS performance evaluation demonstrate a peak cladding temperature not in excess of 2107'F, a peak local cl'adding oxidation percentage. of 7.62K, and a peak core-wide cladding oxidation percentage of less than 0.7%, compared to the ECCS acceptance criteria of 2200'F, 17K, and 1%, respectively. These results were based on an initial SIT pressure of 200 psig, a condition already more adverse for ECCS performance than the 500 psig SIT minimum pressure analyzed for SBLOCA in Section 2.0. Therefore, the Cycle 3 LBLOCA results apply conservatively to Cycle 6 with a minimum SIT operating pressure of 500 psig. In conclusion, operation of St. Lucie Unit 2 Cycle 6 with a minimum SIT pressure of 500 psig at a core power level of 2754 NWt (102K of 2700 HWt) and a peak linear heat generation rate of 13.0 kw/ft is in conformance with 10CFR50.46.
48
Table 3.0-1 St. Lucie Unit 2 Large Break ECCS Performance Analysis Significant Parameters and Initial Conditions
~Paramete alues Core Power Level 2754 at 102% of Nominal (HWt)
Core Average Linear Heat Rate 4.90 at 102K of Nominal (kw/ft)
Peak Linear Heat Generation Rate (PLHGR) 13.0 Hot Assembly, Hot Channel (kw/ft)
Peak Linear, Heat Generation Rate (PLHGR) 11.57 Hot Assembly, Average Channel (kw/ft)
Core Inlet Temperature ('F) 552.0 Core Outlet Temperature ('F) 603.8 System Flow Rate (ibm/hr) 136.lx10 (1)
Core Flow Rate (ibm/hr) 131.lxl0 Gap Conductance at PLHGR' (Btu/hr-ft -'F) 1460 Fuel Centerline Temperature at PLHGR' ('F) 3296 Fuel Average Temperature at PLHGR' ('F) 2102 Hot Rod Gas Pressure (psia)' '118 a
Hot Rod Burnup (NWD/HTU) 1038 Number of Tubes Plugged Per Steam Generator 1430 Augmentation Factor 1.00 Safety Injection Tank (SIT) Gas Pressure (psig) 200 (3)
Initial Containment Temperature ('F) 90 (1) System Flowrate consistent with 363,000 gpm (2) STRIKIN-II values at hot rod burnup which yield highest PCT (3) SBLOCA and non-LOCA use SIT Gas Pressure of 500-650 psig 49
t>>
Table 3.0-2 St. Lucie Unit 2 Results for Limiting Break Size (0.6 DfG/PD)
~Paramete Value Peak Cladding Temperature ('F) < 2107 Time of Peak Cladding Temperature (seconds) 266 Time of Cladding Rupture (seconds) 44.74 Peak Local Cladding Oxidation (5) 7.62 Total Core-Mide Cladding Oxidation P) < 0.70 50
f" l h.
l I
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This section presents the assessment of other Design Basis Events (DBEs) and the station blackout event regarding the reduction of the SIT minimum operating pressure setpoint.
Section 4. 1 reviews the non-LOCA DBEs to determine whether the SIT inventory was credited in any of these events for the St. Lucie Unit 2 safety analyses of record.
Section 4.2 assesses the station blackout event. Station blackout is outside the design basis for St. Lucie Unit 2 but is part of the licensing basis. The station blackout event postulates a four hour time peri'od for restoring AC power during which primary subcooling and natural circulation must be maintained. The station blackout event does credit SIT inventory for maintaining natural circulation.
- 4. 1 Non-LOCA Desi n Basis Events The non-LOCA Design Basis Events (DBEs) were reviewed to determine whether the Safety Injection Tank inventory was credited in the St. Lucie Unit 2 safety analyses of record. In addition, the impact of decreasing the SIT gas pressure on the results of the non-LOCA events was evaluated. The SITs will inject during sever e depressur ization events, releasing highly borated liquid inventory into the RCS. The borated liquid inserts negative reactivity in the core, enhancing the shutdown margin and mitigating a possible return to criticality. Decreasing the SIT pressure (actuation setpoint) would delay the possible introduction of the borated inventory into the RCS and could reduce the margin to criticality.
Table 4. 1-1 presents a list of the St. Lucie Unit 2 non-LOCA DBEs. DBEs which result in a severe depressurization are identified in this Table by arrows in the left margin. These are events during which the reactor coolant system (RCS) pressure decreases below the safety injection actuation signal (SIAS) setpoint and the HPSI pumps shutoff head. The non-LOCA DBEs resulting in such a severe RCS depressurization are:
51
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- 1) Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve
- 2) Post-Trip Return to Power for Steam System Piping Failures (Steam Line Breaks)
- 3) Pressurizer Pressure Decrease Events (Inadvertent Opening of the Pressurizer Power Operated Relief Valves - PORVs)
- 4) Small Primary Line Break Outside Containment (Letdown Line Break)
- 5) Steam Generator Tube Rupture Event Of the above DBEs, the NSSS simulation of the Inadvertent Opening of the Pressurizer PORVs was terminated before the RCS pressure stabilizes. This is due to the fact that the post-trip NSSS behavior for this event falls within the spectrum of considered LOCAs and is not of any interest with respect to the DNB criterion of this analysis. As a LOCA, the post-trip NSSS behavior of this event is within the domain addressed in Section 2.0 of this report.
Review of the remaining DBEs was performed to determine whether the calculated minimum RCS pressure is below the current SIT minimum setpoint pressure of 570 psig (585 psia). This survey shows that the minimum RCS pressure during any of the above DBEs is 661 psig (676 psia) which occurs during the Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric lump Valve event. For the Post-Trip Return to Power Hot Zero Power Steam Line Break Event, the minimum RCS pressure at 300 seconds (the end of simulation time) is 668 psig (683 psia) and essentially stable.
The minimum calculated pressures for,these two events are still above the current SIT minimum setpoint pressure of 570 psig, resulting in no delivery of SIT inventory to the RCS. Since none of the non-LOCA events resulted in or credited the delivery of SIT inventory, a decrease in the minimum SIT setpoint pressure will have no adverse impact on the results and conclusions of these events.
52
0 y U In summary, a review of St. Lucie Unit 2 non-LOCA Oesign Basis Events analyses shows that no credit for SIT injection into the RCS was taken by these analyses in order to show acceptable consequences. Therefore, the reduction of the SIT minimum setpoint pressure has no impact on the non-LOCA safety analyses.
53
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TABLE 4.1-1 ST. LUCIE UNIT 2, DESIGN BASIS EVENTS CONSIDERED IN THE CYCLE 2 SAFETY ANALYSIS Increase In Heat Removal By The Secondary System A. Decrease in Feedwater Temperature B. Increase in Feedwater Flow C. Increased Hain Steam Flow
~ D. Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve E
- Steam System Piping Failures
- l. Inside Containment Pre-Trip
- 2. Outside Containment Pre-Trip Power Excursions
- 3. Post-Trip Return to Power
- 2. Decrease In Heat Removal By The Secondary System A. Loss of External Load B. Turbine Trip C. Loss of Condenser Vacuum D. Loss of Normal AC Power E. Loss of Normal Feedwater F.* Feedwater System Pipe Breaks
- 3. Decrease In Reactor Coolant Flowrate A. Partial Loss of Forced Reactor Coolant Flow B. Total Loss of Forced Reactor Coolant Flow C.* Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft
- 4. Reactivity And Power Distribution Anomalies A. Uncontrolled CEA Mithdrawal from a Subcritical or Low Power Condition B. Uncontrolled CEA Mithdrawal at Power '.
C. CEA Drop D. CVCS Malfunction (Inadvertent Boron Dilution)
E. Startup of an Inactive Reactor Coolant System Pump F.* Control Element Assembly Ejection
- 5. Increase In Reactor Coolant System Inventory A. CVCS Malfunction B. Inadvertent Operation of the ECCS During Power Operation
- 6. Decrease In Reactor Coolant System Inventory
~ A. Pressurizer Pressure Decrease Events
~ B. Small Primary Line Break Outside Containment
~ C.* Steam Generator Tube Rupture
- 7. Hiscellaneous A. Asymmetric Steam Generator Events
- Postulat'ed Accidentsdd dd d
54
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4.2 St t o ac out Event An assessment was made of the validity of the station blackout event-documented in Section 15.10 of the FSAR (Reference 9) with regard to the impact of reducing the minimum SIT pressure setpoint. In this assessment, the results presented in Reference 9 are used to demonstrate that even'ith delayed SIT discharge into the RCS resulting from lowering the SIT setpoint to 500 psig sufficient liquid inventory remains in the RCS to prevent voiding in the RCS loop and loss of natural circulation.
Therefore,. this assessment shows that the conclusions presented in Reference 9 regarding maintaining primary natural circulation ud core subcriticality are unchanged with the reduced SIT pressure setpoint of 500 pslg ~
The initial conditions for the reference analysis are for Cycle 1, and therefore do not include steam generator tube plugging and stretch reactor core power level which have occurred since the Cycle 1 analysis. In the reference analysis, credit is taken for operator action to maintain at least 10'F primary system subcooling in the hot legs. For the Cycle 1 analysis, subcooling ensures condensation of. bubbles produced in the core, restricts void formation due to inventory loss to the reactor vessel upper head and pressurizer, and maintains natural circulation. In this assessment, the effects of tube plugging and power upgrade on the calculated mass and energy balances are assumed to be offset by the difference in steam generator secondary operation that would be required to maintain the specified 'hot leg subcooling. That is, assuming operator actions follow the same "no-load" temperature program, the RCS inventory at a particular RCS pressure for the Cycle 1 analysis will be very similar to the RCS inventory at the same .RCS pressure calculated with tube plugging and stretch power. In a revised analysis, the timing or sequence of events would be different from the Cycle 1 analysis due to the dynamic effects of increased power and reduced steam generator heat transfer area, but crediting operator action to maintain the same level of subcooling (i.e., the same temperature) in the hot legs results in a comparable RCS inventory and pressure relationship. Therefore, a revised station 55
ra~
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blackout analysis is not required for the assessment discussed below, which uses the RCS inventory and pressure relationship in the Cycle 1 refer ence analysis to show that voiding in the hot legs would not be calculated to occur with a reduced SIT pressure setpoint.
The Reference 9 station blackout event analysis was performed using the CESEC-III computer code, see Reference 10. This reference analysis credited SIT discharge to the RCS at about 12540 seconds after event initiation, which was based on a SIT pressure setpoint of 568 psig. Using the Reference 9 calculated results, Table 4.2-1 presents the primary coolant masses in the pressurizer, reactor vessel upper head,. and remainder of the RCS for the time period between 12000 and 14000 seconds.
Columns A and 8 of Table 4.2-1 show that the time of 12000 seconds represents a calculated condition in the RCS before the SIT pressure setpoint is reached (RCS pressure of 632.9 psig). Similarly, the time of 14000 seconds is a calculated condition in the RCS when the primary pressure is below the reduced SIT pressure setpoipt of 500 psig (RCS pressure of 496.6 psig). Therefore, the reference analysis indicates that the time delay in SIT discharge resulting from reducing the pressure setpoint is roughly 1460 seconds.
The data in Columns A through H of Table 4.2-1, is taken from the CESEC output edits for the Reference 9 analysis. Column I in Table 4.2-1 is the mass of liquid in the RCS excluding the upper head and the pressurizer.
This is obtained by subtracting the upper head mass (Column G) from total RCS mass without pressurizer mass (Column F). The total RCS mass'(Column J) including the upper head and the pressurizer masses is obtained by adding the values of Columns E (pressurizer mass) and F (mass of RCS plus surge line). The values at the bottom of each column in Table 4.2-1 are the mass differences between 12000 and 14000 seconds.
Table 4.2-1 shows that from 12000 to 14000 seconds, 11824.4 ibm (Column C) was added to the RCS via SIT discharge in the reference analysis. During the same time period, 1118 ibm (Column D) leaked out of the system.
Therefore, based on the difference between inventory added and inventory lost, the net mass addition to the RCS is 10706.4 ibm. However, Table 4.2-1 also shows that CESEC determines for this time period that the total 56
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RCS mass increased by 16899 ibm (last entry in Column J). This discrepancy between the integrated inventory balance and the summation of total RCS inventory (6192.6 ibm in total inventory gain) is a result of the CESEC computer code numerical integration scheme. For conservatism, the lower value of total RCS mass is used in this assessment of the impact of the reduced SIT pressure setpoint.
Table 4.2-1 shows that the'otal mass of pressurizer and upper head inventory at 14000 seconds is 24559 ibm (13670 ibm from Column E plus 10889 ibm from Column G). Reducing this mass by the total integrated SIT discharge of 11824.4 ibm (Column C) and by the 6192.6 ibm which was added by the code numerical scheme, produces 6542 ibm. This is the hypothetical total mass in the pressurizer and the upper head at 14000 seconds without the SIT inventory. Reducing this hypothetical total mass by the mass of steam produces the total mass of liquid in the pressurizer and upper head.
The steam masses in the pressurizer and upper head are determined using the RCS parameters at 12400 seconds in the reference analysis because this is the last set of information not influenced by SIT discharge. This selection of 12400 seconds 'is conservative relative to later times in the reference analysis because the CESEC-III calculation shows a reduction in vapor mass as RCS pressure decreases, as long as subcooling in the hot legs is maintained. Therefore, the mass of steam that exists in the reactor vessel upper head and pressurizer at this time provides a conservative estimate for the amount of steam that woul'd exist when the RCS pressure decreases to the SIT setpoint pressure of 500 psig.
From the CESEC major edit at 12400 seconds, the mass of steam in the pressurizer is 1829.0 ibm. The mass of steam in the reactor vessel upper head at 12400 second is 1339.5 ibm which is based on (1) a total upper head volume of 1165.4 ft, (2) a void fraction of 0.86547 (Column H at 12400 seconds), and (3) a steam specific volume of 0.753 ft /ibm at a saturated pressure of 598.2 psig (Column 8 at 12400 seconds).
Subtracting the steam mass of the pressurizer and upper head (3168.5 ibm) from the total mass of these regions (6542 ibm) shows that approximately 3373.5 ibm of liquid would still remain in these regions at 14000 seconds 57
Ay A 9)
without any SIT inventory entering the RCS. Column B of Table 4.2-1 indicates that with the reduced pressure setpoint of 500 psig, the SITs will have discharged at some time prior to 14000 seconds. Therefore, it is concluded that RCS liquid inventory would still be enough so that some liquid remained in the upper head and the pressurizer. This means no voiding would have been introduced in the RCS loops.
Restricting voiding to the upper head in. the analysis is the result of assuming operator action to maintain at least 10'F subcooling in the hot legs by opening the atmospheric dump valves to reduce the secondary system pressure and temperature. The increased heat removal in the steam generators caused by the larger temperature difference across the steam generator tubes reduces the primary system temperature. Since the size of the void is determined to remain confined by the upper head and pressurizer, natural circulation in the hot legs is maintained. Discharge of borated water from the SITs prevents additional voig growth. Hence, the conclusions of the Reference 9 analysis regarding'maintaining natural circulation in the RCS for the first four hours of the event are unchanged.
It should be noted that the Reference 9 calculation did not credit the boron contents of the SITs for the reactivity calculations. Therefore, the conclusions of the Refer ence 9 analysis with respect to maintaining core subcriticality are not affected by the SITs pressure setpoint change in this evaluation. Also, this evaluation conservatively used the lower RCS inventory calculation from CESEC and assumed no SIT discharge through 14000 seconds, where the primary pressure is below the SIT pressure setpoint.
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TABLE 4.2-1 STATION BLACKOUT EVENT INVENTORY
SUMMARY
PRESS INTEGRATED INTEGRATED CESEC THERMAL-HYDRAULIC
SUMMARY
RCS w/o UPH RCS TOTAL PSIG SIT,lbm LEAK,lbm PZR RCS Inc UPH UPPER HEAD UPPER HEAD MASS, Ibm MASS, Ibm TIME MASS,lbm MASS,lbm MASS,Ibm VOID FRACT 12000 632.9 0.0 10215A 7.5710E+03 4.4990E+05 8.5076E+03 8.7677E-01 441392.4 457471 12200 622,3 0.0 10332.8 8.6070E+03 4.4880E+05 9.0452E+03 8.6696E-01 439754.8 457407 12400 598.2 0.0 10445.4 8.3660E+03 4.4890E+05 9.1083 E+03 8.6547E-01 439791.7 457266 12600 559.1 2352.1 10560.8 4.8990E+03 4.5580E+05 8.3772E+03 8.7782E-01 447422.8 460699 12800 549.7 3707.9 10678 6.0010E+03 4.5660Et05 8.7643 E+03 8.7086E-01 447835.7 462601 13000 544.5 4411.1 10790.2 7.7420E+03 4.5570Et05 9.2671E+03 8.6196E-01 446432.9 463442 13200 539.1 5226.5 10899A 9.3640Et03 4.5510E+05 9.7185 E+03 8.5398 E-01 445381.5 464464 13400 531.5 6259.9 11005.2 1.0760E+04 4.5520E+05 1.0093 E+04 8.4734 E-01 445107 485960 13600 508.1 9964.3 11115.6 1.0350E<04 4.6120E+05 1.0127Et04 8.4657E-01 451073 471550 13800 501.2 11092.6 11226.6 1.1840E+04 4.6140E>05 1.0396E+04 8.41 85E-01 451004 473240 14000 496.6 11824.4 11333.4 1.3670E+04 4.6070E+05 1.0889 E+04 8.3328E-01 449811 474370 DIFF 11824.4 1118 6.0990E+03 . 1.0800E+04 2.3814E+03 8418.6 16899 DESCRIPTION OF COLUMN HEADINGS A) TIME (SECONDS)
C) INTEGRATED SIT DISCHARGE (LBM)
D) INTEGRATED LEAK FROM RCS, TOTALED FOR THE TWO SIMULATEDLEAKS (LBM)
E) TOTAL PRESSURIZER MASS, PRESSURIZER LIQUID PLUS STEAM MASS (LBM)
F) MASS OF RCS PLUS SURGE LINE (LBM)
G) MASS IN THE UPPER HEAD(LBM)
H) UPPER HEAD VOID FRACTION I) MASS IN THE RCS LOOP, EXCLUDING MASS OF PRESSURIZER AND UPPER HEAD (LBM)
OBTAINED BY SUBTRACTING COLUMN G FROM F J) TOTAL RCS MASS, INCLUDINGTHE PRESSURIZER AND UPPER HEAD MASSES (LBM)
OBTAINED BY ADDING COLUMNS E AND F DIFF: THE LAST ROW PROVIDES THE MASS DIFFERENCE BETWEEN 14000 AND 12000 SECONDS FOR EACH PARAMETER
0 C
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This'eport provides the technical basis for the engineering evaluation which demonstrates that a reduction of SIT pressure minimum operating setpoint from 570 psig to 500 psig for St. Lucie Unit 2 Cycle 6 continues to meet the acceptance criteria for ECCS performance as defined by IOCFR50.46.
This report includes the supporting information and analytical results for SBLOCA, LBLOCA, non-LOCA design basis transients and the station blackout event. The containment peak pressure analysis was not examined by this study, however, the reduction of SIT pressure minimum operating setpoint to 500 psig does not adversely affect the containment peak pressure because the total mass and energy released to the containment from the primary system in the analysis is not changed and because the SIT discharges less than one second later at the lower setpoint.
Reducing the technical specification LCO value to 500 psig provides additional differential pressure margin between the SIT operating pressure and the SIT relief valve pressure setpoint (669 psig). This additional margin lessens the potential for challenges to'he SIT relief valve and concurrent relief valve leakage that have impacted St. Lucie Unit 2 plant availability.,
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CENP0-137, "Calculative Methods for the CE Small Break LOCA Evalu'ation Model," Combustion Engineering Proprietary Report, August 1974, (Proprietary).
CENP0-137, "Calculative Methods. for the CE Small Break LOCA Evaluation Model," Supplement 1, January 1977, (Proprietary).
Acceptance Criteria for Emergency Core Cooling System for Light-Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No.
3 -Friday, January 4, 1974.
Letter, E. L. Trapp (C-E) to W. L. Parks,(FPLL), "St. Lucie Unit 2.
Cycle 6 Reload Safety Evaluation (RSE) Report," F2-90-035, July 6, 1990.
CENPD-133, Supplement 1, "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident," August 1974, (Proprietary).
CENP0-133, Supplement 3, "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident," January 1977, (Proprietary).
CENP0-134, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," April 1974, (Proprietary).
CENPD-134, Supplement 1, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modification)," December 1974, (Proprietary).
CENPD-134, Supplement 2, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," June 1985, (Proprietary).
CENPD-135, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1974, (Proprietary).
CENP0-135, Supplement 2, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modification)," February 1975 (Proprietary).
CENP0-135, Supplement 4, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," August 1976, (Proprietary).
CENP0-135, Supplement 5, "STRIKIN-II, A Cylindrical Geometry'uel Rod Heat Transfer Program," April 1977, (Proprietary).
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- 7. CENPD-138, "PARCH - A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," August 1974, (Proprietary).
CENPD-138, Supplement I, "PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup,"
(Hodification), February 1975, (Proprietary).
CENPD-138, Supplement 2-P, January 1977, (Proprietary).
- 8. Letter, E. L. Trapp (C-E) to J. L. Perryman (FPKL), "Results of CE's Large Break LOCA, Reevaluation for St. Lucie Unit 2 Based on CE's New (1986) Evaluation Model," F2-CE-R-137, April 10, 1987.
- 9. FPEL, "St. Lucie Plant Unit 2, FSAR," Amendment No. 1, April 1986.
- 10. C-E Topical Report, CENPD-107, "CESEC - Digital Simulation of a Combustion Eng'ineering Nuclear Steam Supply System," April 1974.
C-E Topical Report, "CESEC - Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," Enclosure 1-P to LD 001, January 6, 1982.
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