ML17266A494
| ML17266A494 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 09/08/1981 |
| From: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| L-81-389, NUDOCS 8109090226 | |
| Download: ML17266A494 (108) | |
Text
REGULATORY DEFORMATION DISTRIBUTION 8 'EM (R IDS)
ACCESSIONS NBR:81090902'26'OC DATE'1/09/08 ~ NOTARIZED!'O DOCKEiT' FACIL~ 50 389 St'>> Lucie Plant~ Unit 2~ Flor ida Power ((i Light Co>> 05000389 AUTH", NAME'UTHOR AFF ILI'ATiION UHRIGg R, E ~ Florida Power (( Light Co ~
BECIP ~ NAMKI RECIPIENT AFFIL'IATION EISENHUT'iD>>G, Division of Licensing
SUBJECT:
" For wards response'o NRC requests for addi info which have not be'en formally submitted on docket. Responses will in future~ emend to FSAR, oe'ncorporate'd DISTRIBUTION CODKit BOO'1S COPIES RECE{IVED:L7R 'NCLt 'IZE:;i i T'ITLEl PSAR/FSAR AMDTS and Re.lated Correspondence NOTES!
RECIPIENT COPIES RKC IP IENT"'D COPIES ID CODE'/NAME{ LTTR ENCL.-'- CODE'/NAME: LTTR{ ENCL ACTION:= A/D LICENSNG 1 0 LCC BR 43 BC 0 I.IC BR ¹3 LA 1 0 NERSESgV ~ 04 1 1 INTERNAL: ACCID EVAL BR26 1 1 AUX SYS BR 2'7 1 1 CHEM ENG BR 11 1 1 CONT SYS BR 09 1 1 CORE PERF BR 10, 1 1 EFF TR SYS 1 li PRP LIC BR12'MRG EMRG PRP DEV 35i 1- 1 3H EQUIP'UAL BR43i 3i 3 RKP DIV 39 36'EMA 1 1 GEOSCIENCES 28 2 2 HUM FACT ENG 40 1.
HYD'/GEO BR 30. 2 2 I((C, SYS" BR 16 1 1 IE Ki 06' 3 3 LCC GUID BR 35 1 1 IC QUAL BR 32". 1 'i{ATL'NG BR 17 1 1 MECH ENG BR 18 1 MPA 1 0 OELD 1 0 OP. LIC BR 34 1 1 POAER SYS'R 19 1 1 PROC/TST REV 20 1 1 QA BR 21 1 1 R'AD SS BR22'1>> 1 1 RE'AC SYS BR 23>> 1 1 1 1 SIT'NAL'R 24 1 1 RUCT ENG- BR25 1 1 EXTERNALi: ACRS 41i ib 16 LPDR 03 1 1 NRC PDR'TIS 02i 1 1 NSIC 05 1 1 1 1 8FF g y g~,
TOTAL. NUMBER'~ OF COPIES REQUIRED: LiTTR 62 ENCL~ 57
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/Isv<4 FLORIDA POWER st LIGHT COMPANY September 8, 1981 Letter L-81-389 Office of Nuclear Reactor Regulation Attention: Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear Mr. Eisenhut:
Re: St. Lucie Unit 2 Docket No. 50-389 Final Safety Analysis Report Re uests For Additional Information Attached are Florida Power 6 Light Company (FPL) responses to NRC staff requests for additional information which have not been formally submitted on the St. Lucie Unit 2 docket. These responses will be incorporated into the St. Lucie Unit 2 FSAR in a future amendment.
Very truly yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/TCG/nlc Attachments cc: J. P. O'Reilly, Director, Region II (w/o attachments)
Harold F. Reis, Esquire (w/o attachments)
((0 t gi'I 8f0909022b 8i0908 PDR ADOCK 05000389 A PDR HELPING 8UILD FLORIDA
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.Attachments to L-81-389 September 8, 1981 A. ~
Minutes to Structural Engineering Branch meeting held on September 3, 1981.
B. Revision to Reg. Guide 1.97 Rev. 2 (Additional information on Steam Generator water level indication).
C. Revised response to question 460,.2 D. Revised section 9.3.6 (Post accident sampling).
E. Revised section 15.1.5 (Limiting fault 3 events).
F. Revision to FWLB Analysis in response to informal request on AFW flow requirements.
G. Response to question 252.1 H. Revised response to question 420.05 Additional information to existing response on SER 8.3 open item 90%
480V motors.
J. Chapter 8.3 open item Isolation devices.
8109090226
A meeting was held between the NRC, Structural Engineering Branch, Flordia Power & Light Co. and Ebasco Services on September 3, 1981 in Bethesda, MD. Those in attendance were:
NRC FP&L Ebaseo V. Nerses W.F. Brannen W. Fan P.T. Kuo E.W. Dotson E. Kowalski H.E. Polk R. Russo L. Gertler J. Burket P. Grossman Items discussed are indicated below:
Audit Item 2 FP&L resubmitted a revised response, attached.
The NRC has agreed to close this item pending staff review.
Audit Item 3 FPL has agreed to provide the NRC a completed re-sponse by September 9, 1981. This is considered an open item.
Audit Items 4 & 5 FP&L has submitted clarification of headings on the Reactor base mat shear stresss table, attached.
In addition, FPL will provide the water ford para-metric study detailing the sensitivity of the base slab structural stresses to the soil thickness used in the finite element analysis by September 9, 1981.
Item is considered open pending NRC review.
Audit Item 7 FP&L has agreed to resubmit a revised response,.
attached. The NRC has agreed to close this item pending staff review.
Audit Item 8 FP&L will resubmit Figure 9 to indicate radial and
.tangential components of seismic loading. by September 9, 1981. A legend will be provided to explain the variables shown on the computer printouts. Resubmittal will close this item.
Audit Item 11 NRC is reviewing the response to this item. Item closed.
Audit Item 14 FP&L will provide mode shapes for horizontal ac-celeration with Kxx and Kxx=0. This will be provided
. for information only. This item is considered close.
Audit Item 15 FP&L resubmitted a revised response referencing audit:item 18 for additional clarifications. Item considered open.
Audit Item 16 NRC is reviewing the use of SRSS summation vs.
absolute summation regarding piping relative dis-placements between structures. This item still considered open.
Audit Item 17 FP&L will provide applicable portions of the CB&I stress report and the magnitude of the stresses for pipe attached to the containment shell by September 9, 1981. Item closed pending NRC review.
Audit'tem 18 FP&L provided a response to this item, attached.
This item considered open pending NRC review.
Audit Item 19 FP&L resubmitted a revised response adding a reference to page 46, attached, of typical containment building calculations, Section 3. This item considered closed.
Audit Item 20 FP&L resubmitted a response justifying the use of the 0.63 factor. This item considered open pending NRC review.
Audit Item 21 FP&L will provide mode shapes to the NRC and deter-mine the amount of shear area reduction by September 9, 1981. This item will be closed pending NRC review.
Audit Item 22 FP&L will verify that the wall section investigated is the most critical. In addition, FPL will verify the OBE/DBE ratio of 0.66 was used for the design of the reinforcing steel by September 9, 1981. Favorable NRC review will close this item.
Audit Item 24 FP&L provided a response to this tiem, attached. Item considered open pending NRC review.
A'udit Item 25 FP&L provided a response to this item, attached. Under NRC review (see audit item no. 16).
Audit Item,26 FP&L provided a response to this item, attached. Item is considered closed.
Masonry Wall Criteria FP&L has agreed to'tilize the NRC's criteria for masonry block wall design dated 'July 1981. This item is considered closed.
NRC Question 220.22 FP&L provided a response to this question, attached. Items is considered open.
NRC Question 220.25 FP&L provided a response to this question, attached. Items considered open pending NRC.
Turbine Missile FP&L will provide an explanatian of the formula used for barrier yene'tration. by September 9, 1981.
Favorable review by NRC will close item.
I I gt Item No. 2 Verify compatibility of boundary conditions between inter-facing structures analyzed by different computer programs.
~Res oese Compatibility of the two computer programs was not considered in the original analysis because the design of the structure was based on the most conservative boundary assumption. (Fixed end condition for negative moments and hinged for positive moments.
The analysis of the secondary shield wall was done assuming fixed boundary at the connection with the primary shield wall h
The resultant boundary reactions derived from the analysis of the secondary shield wall are applied as additional loading
. to,the upper boundary of the primary shield wall.
See attached Fig. 2-1 and calculations.
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R.G. 1.97 Rev. 2 y0 VARIABLE EXISTING RANGE REQUIRED RANGE CONIENTS TAG NO. DESCRIPTION PIA-1116 Quench Task D15 0 to design pressure 0-100 paig One non-safety pressure indicator is pro-Pressure vided in the control room. Design pre-sure is 100 paig therefore the range provided is acceptable.
LIC-9013A From tube sheet to separa-LIC-9013B Steam Generator 916 Redundant safety. grade narrow range in-Level tors See Figure dicators are provided in 'the control room LIC-9013C 420.41-1 LIC-9013D with a span of 183" (LIC;9013A,B,C,D and LR-9013D.. LIC-902:3A,B,C,D)'. One channel per steam LIC-9023A generator is also recorded in the control LIC-9023B room (LR-9013D, LR-9023D). In addition wide LIC-9023C range non-safety indication for each opera-LIC-9023D tor is provided in the control room with a LR-9023D. span of 465" (LI-9012, LI-9022). The upper LI-9012 tap for wide and narrow range is located at LI-9022 .the same level and is approximately 40" abov the steam separators. The narrow range is located approximately 57" below the uppermos steam generator tubes and 50" above the low-est steam generator tubes. The lower tap fo the wide range level is located 465" below the upper tap. The lower tap is approximatel 20" above the tube sheet. The combination of these instruments cover the range required although not all the instruments are safety related.
PI-8013A Steam Generator D17 From atomospheric pressure Redundant safety grade indication is pro-PI-8013B Pressure to 20X above 'the lowest vided in the control room. One channel:
PI-8013C safety valve setting. per steam generator is recorded. The PI-8013D range provided is not adequate'and will PR-8013D be changed to meet the requirem'ents to PI-8023A 0-1200 psia.
PI-8023B PI-8023C PI-8023D PR-8023D
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VARIABLE EXISTING TAG NO. DESCRIPTION RANGE REQUIRED RANGE COMMENTS FI-08-lA Safety/Relief D18 Closed Not Closed Main Steam flow is indicated in the CR.
FI-08-1B Valve Positions Flow This measurement is accepted for Safety Relic valve position indication. Qualified sensors are being utilized.
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Question No.
460.2 (11.2) List outdoor storage tanks that may contain po-tentially radioactive liquid and describe pro-visions designed to prevent, collect, and process spills, from outdoor storage tanks. It is our position that outdovr tanks should be designed in accordance, with Regulatory Position 1.2 in Regulatory Guide 1.143, Rev. 1.
Res onse:
'ank Refueling Water Tank (RWT)> Condensate Storage The (CST), Primary Water Storage Tank (PWST) and the Steam Generator Blowdown Monitor Tank (SGBMT) are outdoor storage tanks which could contain potentially radioactive liquid.
The RWT, CST and PWST are provided with control room level indication and high level alarm. Thy le~<~
CST and PWST are also provided with local high~alarm and the SGBMT is provided local level indication and high level alarm.
The SGBMT overflow and drains are routed to the equipment drain tank in the Liquid Waste Management System. The CST is surrounded by the CST building so that overflow and leakage are collected in the building. Level control systems will be provided for the RWT and PWST, to prevent tank overflow.
FSAR Subsection 6.3.2.2.4 and 9.2.3.1.5 will be revised to incorporate the RWT level control system and the'WST level control system respectively. The RWT and'WST level control system will be implemented as a backfit item after OL date.
RWT LEVEL MONITORING SYSTEM Ebasco has reviewed .the St Lucie Unit No. 2 piping design to overflow'occurence of the possible due to the ar-determine'hether RNT may be rangement of the RWT volumes and lf 4 j fill lines.
The SL-2 RWT is a 525,000 gallon capacity, 40 feet. high, atmospheric tank used to hold=borated water for refueling and safety injection=-
operations. The RWT is normally maintained at a minimum tank level of 28.2 feet for'safety injection purposes (413,600 gallons required) and at a maximum tank level of 35.3 feet for refueling purposes (Approximately 500,000 gallons). The RWT is provided with an over-flow level at 38 feet. and a high level alarm set at 37.5 feet, thereby providing 7,520 gallons before tank overflow. The present, RWT level monitoring design, depicting the various RWT elevations, is provided schamatically on attache Figure 460.2-1 while a summary of the RWT fill lines is provided in Table 460.2-1.
We have evaluated the possibility of RWT overflow from each of the fill lines and have tabularized the overflow time for each source (see T~6(e l&o. 2- ) . It was determined that- the limiting RWT fill I
2.4 minutes to respond before the tank overflow occurs.
tank fill 'll source is the 3,100 GPM from the LPSI pumps which allows the operator sources provide the operator a minimum of twenty (20) other minutes before'ny a'ction is required.
Ebasco has developed an automatic RWT isolation design which will provide enhancements to,the current. design which will increase the operator response time margin and minimize the potential for RWT overflow.
This level monitoring system will include an air-operated, fail-closed automatic block valve to the LPSI line, which will close upon receipt of high RWT level signal. The existing high RWT level al'arm is retained and additional annunciation of high-high RWT level at 37.75 feet, which provides 3760 gallons below overflow, will be provided. This annunciation will receive a signal from a source different from the present high-level alarm signal for ad-
. ditional reliability and will be alarmed in the control room. This alarm will give the operator additional time to initiate any positive action if the-RWT fill is inadvertently left open.
1 I PNST LEVEL CONTROL SYSTEM The flow of water into the primary water tank is controlled by a level, control valve in the, intake header (demineralized water supply) to the tank. High and low water levels in the tank are annunciated.
In addition to the above, interlocks will be provided to auto-matically stop the waste condensate pumps and boric acid con-densate pumps which discharge to the PWST (see Figure 9.2-4)
'upong high water level in the tank.
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RATED FLOP HIGH ALARM TO LINES TO RWT SOURCE ( m) OVERFLOW (MIN) 6"-CS-500 LPSI Pumps 3,100 2.4 6"-SI-154 HPSI Recirc. 30 250 LPSI Recirc aoo 75 Recirc 'S 150 50 3"-CS-62 Reactor Drain Pumps 50 150 3"-NN-A56 Hold Up Drain 6 Recir- 80 94 culation Pumps 3"-PM'-16 Primary Hater Pumps '325 .23 3"-FS-556 Fuel Pool Purification Pump 150 50 I-3-CH-938 Boric Acid Pumps 150 50 F
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9.3.6 POST ACCIDENT SAMPLING SYSTEM The Post Accident Sampling System (PASS) consists of a shielded skid-mounted sample station and a remotely located control panel.
The PASS provides a means to obtain and analyze pressurized and unpressurizeg reactor coolant samples and containment building samples.
The Piping and Instrumentation diagrams for the PASS are shown in Figures'9 3-7 go 9'5-V'. Design data is provided in Tables 9.3.10, 9.3.11 and 9.3.12.
9.3.6.1 ~0i B The PASS is designed in accordance with the criteria stated in Section II.B.3 of Enclosure 3 to NUREG 0737. The quantitative design criteria for the PASS are as follows:
a) The PASS provides a means to promptly obtain a reactor coolant liquid, containment building sump liquid, and containment build-ing gas samples. The combined time required for sampling and analysis is less than three hours. .
b) The PASS allows for post-accident sampling with resulting per-sonnel radiation exposure not exceeding the criteria of GDC 19 (Appendix A to 10 CFR Part 50).
c) The PASS is capable of accomnodating an initial reactor coolant radiochemistry spectrum corresponding to a postulated release equivalent to that assumed in Regulatory Guide 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, Rev. 2 dated June, 1974, and Regulatory Guide 1.7,, Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Ac-cident, Rev.' dated September 1976.
d) The PASS provides a means to remotely quantify pH and the concen-trations of total dissolved gas, hydrogen, oxygen and boron in the liquid samples.
e) :Sample flow is returned to the containment to preclude un-necessary contamination of other auxiliary systems and to ensureI that radioactive waste remains isolated within the lt containment.
f) Components and piping are designed to guality Group D (as defined in Regul a tor y 'ui de 1. 26) non- sei smi c requirements.
The equipment is located downstream of double isolation valves from safety code systems.
9.3.6. 2 S s tern Des cri ti on The requirements for post-accident sampling of the reactor coolant and containment building atmosphere are met through the Post-Acci-dent- Sampling System (PASS). The PASS provides a means to obtain pressurized and unpressurized reactor coolantsamples and containment building atmosphere samples. A reactor coolant sample can directly from the Reactor Coolant System (RCS) whenever be'rawn the RCS pressure is between 200 psig and 2485 psig. RCS sample lines are provided with orifices'nside containment so as to limit, the flow from any postulated break'in the sample line.
At pressures below 200 psig, reactor coolant samples can be drawn from a Safeguards System sample line. Tliis pathway also provides a means of sampling the containment building sump during the recir'culation mode of Safeguards System operation. A containment building atmosphere sample can be drawn with containment building pressure between 10 psia and 75 psia. All sample flow is returned to the containment building to preclude unnecessary contamination of other auxiliary systems and to ensure that high level waste re-mains isolated within the containment. These sample process path-ways were selected to insure a representative sample under all modes of decay heat removal. The PASS sampling flow rates are provided in Table 9.3. 10.
0 The PASS consists of a remotely located control panel and a skid-mounted sample station which are designed to maintain radiation exposures to plant personnel as low as reasonably achievable (ALARA) and which is located to minimize the length of sample lines. The PASS is interfaced with the existing reactor coolant and safeguards system sample lines.
Post accident sampling does not require an isolated auxiliary system to be placed in operation.
The PASS is a totally closed system (i.e., samples taken from containment are returned to the containment). The grab samples are extracted from sample vessels by injection of a syringe
. through a septum plug mounted in the vessels. In addition, the PASS sample station skid is provided with a ventilation flowpath that is sized for 333 scfm in air flow from the surrounding room to the ventilation system exhaust. The exhaust air is directed through an activated charcoal filter for iodine removal.
The PASS provides the capability for remote chemical analyses of the reactor coolant including total dissolved gas concen-tration, dissolved hydrogen and oxygen concentration, boron concentration and pH. Reactor coolant analysis is provided through the use of an undiluted grab sample facility.
Shielded grab samples of the depressurized undiluted reactor coolant liquid may be obtained. Unshielded, depressurized and diluted grab samples of the degassed reactor coolant liquid, reactor coolant dissolved gas and containment building atmosphere may also be obtained.
The operation of the PASS for collecting and analyzing reactor coolant and containment building atmosphere samples may be categorized as (1) reactor coolant sample purging, (2) reactor
coolant sample gaseous analyses and dilution, (4) undiluted liquid grab sample col'.ection, (5) containment building .
atmosphere sample purging and dilution, and (6) system flush-ing. An operation description for these categories is provided below:
Reactor coolant sample purging is accomplised by directing the sample flow through the system isolation valves, the sample vessel/heat exchanger, the pressure reducing throttle valve, and out to the containment building. sump. At reactor coolant pressures of less than 2QO psig the containment sump sample flow is purged in the same manner using the safeguards pump discharge connection.
Reactor coolant gaseous analysis is performed on a pressurized sample which is collected by isolating the sample vessel/heat exchanger. Total dissolved gas concentration is determined by degassing the sample. This is accomplished by depressurization and .circulation, by alternate operation of the burette isolation-valve and the sample circulation pump. The resulting displace-ment of liquid into the burette is used to calculate the dis-gas'oncentration. The collected gases, which have been 'olved stripped .from the liquid, are then directed through a float valve for moisture separation and circulated through hydrogen and oxygen analyzers. After recording the hydrogen and oxygen gas concentrations, the gas sample vessel, which contains nitrogen, may be 'placed on line to dilute the gas volume. This dilution operation reduces the radiation, levels such that local samples can be drawn from the gas sample vessel, if desired, by injec-tion of a syringe through a septum plug mounted in the vessel.
Prior to sample withdrawal, additional dilution, which may be for this quantification, may be performed by further 'ecessary nitrogen addition, circulation and venting.
Reactor coolant liquid analyses is accomplished by reinitiating and directing the sample flow through the in-line chemistry analysis equipment. The gas residence chamber and float valve downstream of the throttle valve allows for automatic venting of gases coming out of solution. This venting is required in or'der to prevent gas bubble interference with flow rate and chemistry measurements in the downstream instrumentation.
Boron and pH readings are obtained from the in-line instrumen-tation. A small fixed volume of depressurized liquid sample (collected in a four-way valve) is then drained to the depres-surized liquid sample vessel and a sample is withdrawn in the same manner as described above for the gas sample.
An undiluted liquid grab sample for chloride analysis can be collected by directing reactor coolant purge flow through the undiluted depressurized liquid sample vessel. This vessel is provided with a lead shielded container and cart for transfer of sample to the analysis location. The isolation valves for the vessel are provided with stem extensions penetrating the shielding.
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+ Containment building atmosphere sampling is initiated by-'opening the containment isolation valves and by using the containment sample pump to purge the air sample through the system. Purge flow is directed back to containment. A sample is'manually withdrawn from the containment sample vessel contain-ing nitrogen. The initial nitrogen volume dilutes the sample to levels acceptable for withdrawal. A containment air sample may then be withdrawn from the containment sample vessel in the same manner as described previously for the reactor coolant samples.
System flushing of the liquid and gaseous portions is accom-plised by purging with demineralized water and nitrogen, respec-tively, to reduce personnel exposure during withdrawal of the diluted samples and to reduce contamination plateout between samples.
'x f Radionuclide analyses are performed on grab samples. These samples are counted in standard radionuclide counting equipment.
Grab sample techniques are utilized for 4~4~ analysis.
Backup boron analysis is performed using atomic absorption techniques.
Containment hydrogen analyzers are described in Subsection 6.2.5.
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g onent Descri tion
'.3.6.
3 Com The major PASS components are described in this section. The principal component data summary including design code is pro-vided, in .Table 9.3.11.
~51 St ti The sample station is a free-standing skid-mounted enclosure.
The enclosure contains the piping, valves, components and in-strumentation necessary to provide the sampling and analysis capability. The enclosure is provided with louvers sized to pass up to 333 scfm from the surrounding room to the ventila-tion system suction connection in the upper portion of'.the enclosure. This air flow precludes any possible buildup of radioactive or hydrogen gas and provides for removal of heat generated by internal components. The enclosure is provided wi.th removable panels on all four sides to ensure accessibility for maintenance.
- 2. 'Sam le Circulation Pum The'ample circulation pump is a peristaltic type postitive di'splacement pump. This pump is capable of pumping liquids and/or gases. The pump will be used in the total gas, hydrogen, and oxygen gas analyses operations to strip the gases out of solution in the sample fluid and circulate them through the hydrogen and oxygen analyzers.
- 3. Sur e Vessel Pum The surge vessel pump is a progressing cavity (helical) pump. The pump is used to pump down the surge vessel con-tents to the containment building sump and is also used in the calibration operation of the pH in the liquid sample line.
- 4. Containment Sam le Pum The containment sample pump is a vacuum pump/compressor unit that operates as a positive displacement compressor using a stainless steel diaphram.'he pump is used to collect a containment atmosphere sample and to dilute the sample via circulation through the containment sample vessel.
- 5. Gas Sam le Vessel The gas sample vessel is a 12,000 ml sample vessel initially filled with nitrogen gas. The vessel supplies the gas analysis loop with nitrogen gas to dilute the radioactive gases present in the sample line. The 'vessel is equipped with a septum plug which allows the operator to withdraw a diluted gaseous sample with a syringe for radiological analysis.
- 6. De ressurized Li uid Sam le Vessel The depressurized, liquid sample vessel is a 12,000 ml sample vesse'i. This vessel collects a liquid sample trapped in the four-way valve located above the sample vessel. The vessel is partially filled with demineralized water before the sample is drained into the vessel. Additional demineralized water is then added to obtain the proper dilution factor so that a liquid sample can be withdrawn for radiological analysis. This =vessel is equip-ped with a septum plug for sample withdrawal using a syringe.
- 7. Containment Sam le Vessel The containment sample vessel is a 12,000 ml sample vessel that'is initially filled with nitrogen gas for dilution.
The containment sample pump draws a sample from containment and circulates it through the sample vessel where the nitrogen gas dilutes the sample so that it can be withdrawn for radio-logical analysis. This vessel is equipped with a septum plug for sample withdrawal.
- s. ~s' The surge vessel has a 10 gallon capacity and serves as a vent and drain tank for the depressurized liquid sample vessel and the total gas analysis burette. This vessel can also be filled with buffer solution used to calibrate the in-line pH meter.
Sam le Vessel/Heat Exchan er The 'sample vessel/heat exchanger, is a vertically mounted, shell and tube type heat exchanger. The heat. exchanger uses component cooling water to cool the reactor coolant sample flow from a maximum RCS temperature of 650o to 120oF to allow low temperature sample analysis. The tube side of the heat exchanger serves as as sample vessel for collection of a pressurized reactor coolant sample.
- 10. Stainless Steel Burette The stainless. steel burette has a 1,000 ml capacity. The burette is used to determine the amount of total gas present in the sample fluid by measuring a difference in the -fluid level of the burette upon degassification'f the pressurized reactor coolant sample.
ll. Strainer The strainer is designed to remove insoluble particles which may cause sample station chemistry instrumentation to become plugged. The strainer can be backflushed with demineralized water remotely by operation of valves at the control panel.
- 12. Grab Sam le Facilit The grab sample facility is designed to obtain a 75 cc undiluted sample of reactor coolant liquid. The facility consisted of a lead shielded sample vessel and valves mounted on a cart for transport within the plant. The facility is manually operated.
- 13. Gas Residence Chamber The, gas residence chamber is a horizontally mounted lead shielded baffled cylindrical vessel. The chamber is used to remove undissolved gases from reactor coolant samples to prevent interference with the in-line process monitors.-
- 14. Charcoal Exhaust Filter The charcoal filter is designed to remove radioactive iodine and particulate material from the enclosure ventilation exhaust.
The filter is mounted in a separate housing located on top of the sample skid enclosure.
9.3.6.4 Instrumentation and Control Descri tion The major PASS instruments and controls are described in this section". The on-line process monitor data is provided in Table
-9. 3.12.- g+g Qg 7 P
- 1. Control Panel The panel is designed to meet NEHA-12 requirements. All sample system non-code isolation valves and pumps are con-trolled from this panel.= Indication of all process para-meters and chemistry readouts are displayed on the panel; To facilitate system and operability all controls and indi-cations are arranged in a mimic of the system. All process pumps and valves are equipped with hand switches at the control panel.
The containment building atmosphere sample piping is heat traced to limit plateout of radioiodine and condensation of containment atmosphere vapor. The heat tracing ensures lumeter a representative gas sample.
+ WPsB@r E,
- 3. Boron The Boron Heter is a specific gravity measuring device which determines and remotely indicates the concentration of boron present in the liquid sample.
- 4. ~HHeter The pH meter determines and remotely indicates pH in the liquid sample.
- 5. H dro en Anal zer The hydrogen analyzer is a thermal conductivity device that determines and remotely indicates the volume percent of hy-drogen in the gas stripped from the reactor coolant.
The oxygen analyzer is a paramagnetic device that determines and remotely indicates the volume percent of oxygen in the gas stripped from the reactor coolant.
9.3.6.5 S stem Evaluation The location of the post-accident reactor coolant and containment atmosphere sampling system are in an area of relatively low post-accident background radiation. This ensures compliance with the personnel exposure limits of NUREG 0737 during sampling and analysis. Additional plant shielding along with selective routing of interconnecting piping to the existing sampling system ensures that (1) the exposure limits for personnel are not exceeded and (2) the on-site radiochemistry analysis equipment is available for
t post-accident sample analyses. The sample station is also physical-ly separated from safety related equipment such that failure of the associated
~,,t5 non-seismic equipment does not cause damage to the safety, .related equipment.
S pfSB R.T Coolin'g water to the reactor coolant sampling system is available during post-accident conditions to enable low temperature sample analyses. Overrides are also available to enable opening of con-tainment isolation valves following a CIAS so that post-accident sampl'ing can be accomplished. Control for the reactor coolant sampling-system return containment isolation valve is provided in the, control room. An interlock is provided to ensure that this valve and'he containment sump isolation valve is open before the system, inlet isolation valve is open.
As much as practicable, reactor coolant sampling system connecting piping is pitched downward at least 10 degrees to prevent settling or separation of solids contained by the sample. Traps and pockets in which- condensate or crud may settle are avoided since they may be partially emptied with changes in flow conditions and may.result in sample contamination.
9'."3. 6. 6 Testin 'and Ins ection The sample station skid and control panel are equipped with doors for testing and inspection during normal operations. The sample station is provided with removable panels on all four sides for inspecti'on. Each. component is tested and inspected prior to in-stallation in the sample system. Instruments are calibrated during'nitial system installation. Automatic controls are tested for actuation at the proper setpoints. The system is operated and tested upon installation with regard to flow paths, flow capacity and mechanical operability.
TVSER7 F
Periodic calibration is performed. according to the schedule provided in Table 9.3.13. The. PASS is designed to function for six months'under post-accident conditions without recalibration.'ystem operability will be tested at a frequency minimum of six months, coinciding with the required six-month Emergency'lan sampling exercise; Such operability tests will check the functioning of all aspects of the system.
Xl 9.3.6.7'0 ez.ator Trainin All FP8L Chemistry Departmeht technicians will be trained. both in the class-room and in actual hands-on operations, as a function of the Chemistry training program. Operating procedures will be developed and they will Depart-'ent be consistent with the recommendations of the PASS supplier (Combustion Engineering)
~ Insert A In accordance with item .II.B.3 of NUREG-0737 (pg 3-67,item 4), PASS has the capabil i ty to moni tor total dissolved gases and H2 concentr ati on. The cap-ability"of monitoring dissolved 02 will be in accordance with Regualtory Guide 1.97 (R2).
Insert C Sufficient shielding will be provided around the post-accident sampling system components to limit personnel exposure to .the GDC-19 limits. Regulatory Guide 1.4 source terms will be used.
Insert D The post accident sampling panel will be powered from power panel 2AB which is cap-able of being powered from the diesel generator in the event of a loss of offsite power. The electrical cables associated with the post accident sampling panel and associated '.nstruments will be routed in accordance with Reg. Guide j..75, physical Independence of Electrical System (Rev 1), as associated circuits.
Insert E Heat tracing circuits will'e electrically connected to the respective Boric "Acid Heat Tracing Panels which are electrically independent, physically separated and are connected to the diesel generator in the event of a loss of offsite power.
Insert F Post-accident, sam)ling system valves which are required to operate after an, accident and are not accessible for repair will be qualified to the accident environment in which they operate. Environmental qualification is addressed in FSAR Section 3.11.
nsert B Radiological analysis..of PASS grab samples will be used to identify the occurrence and type of core damage.'Laboratory analysis of PASS grab samples using germanium detectors and multichannel analysis will be employed to identify the presence of selected radiosotopes which are indicative of the various kinds of core damage.
Core damage will be categorized according to clad failure, fuel overheat, and fuel melt. The analysis will take into account the core burnup, coolant water volume, and coolant temperature corrections. A plant procedure in compliance with NUREG-0737 will be written by June 1982.
Table 9.3.10 Post-Accident Sam linq S stem Flow Rates Nominal Source Flow Reactor Coolant Hot Leg 0.2 - 1.0 gpm Containment Buil ding Sump 0.2 - 1.0 gpm Containment Atmosphere 0.2 cfm
Table 9.3.11 Desi n Data for Post-Accident Sam lin S stem Com onents Sam le Circulation Pum Type Peri stal ti c Posi ti ve -Di spl acement Fluid Post-Accident Reactor Coolant Suction Pressure (max) psig 5 Suction Temperature (max) oF 160 Rated Flow, gpm 1 Rated Head, ft 50 Non-Code Code Sur e Vessel Pum Type Posi ti ve Di spl acement Fluid Post-Accident Reactor Coolant Suction Pressure (max) psig 5 Suction Temperature (max) ~F 160 Rated Flow, gpm 1 Rated Head, ft 185 Non-Code Code Containment Sam le Pum Type . Vacuum Pump/Compressor Fluid Post-Accident Containment Atmosphere Suction Pressure (max) psia 10-75 Suction Temperature (max) oF 300 Rated Flow, .cfm 0.2
/
Maximum Discharge Pressure, psig 95 Code Non-Code Sam le Vessel/Heat Exchan er Type Shell (cooling); Tube (sample flow)
Tube Sides:
Fluid Post Accident Reactor Coolant Piping Design Pressure (max) psig 2485 Inlet Temperature (min/max) oF 120/650 Shell Side:
Fluid Component Cooling Mater Piping Design Pressure,'sig 150 Inlet Temperature (min/max) oF 65/120 Flow (max) gpm 30
'ode Non-Code
Table 9.3.11 (cont'd)
- Desi n Data for Post-Accident Sam lin S stem Com onents
- 5. De ressurized.'Li uid Sam le Vessel Internal Volupe, cc 12000 ml Design Pressure, psig 50 Design Temperature, oF 200 Operational Pressure, psig 5 Operational'emperature, F 120 Material Stainless Steel 316L Fluid Post-Accident Reactor Coolant Code - Non-Code
- 6. Gas Sample Vessel Internal Volume, cc 12000 Design Pressure, psig 50 Design Temperature, oF 200 Operational Pressure, psig 5 Operational Temperature,, oF 120 Material Stainless Steel 316L Fluid N2, H2, 02, Fission Products Code Non-Code
- 7. Containment Sam le Vessel Internal Volume, cc 12000 Design Pressure, psig 50 Design Temperature, oF 100 Operational Pressure, psig 0 to 20 Operational .Temperature, oF 275 Material Stainless Steel 316L Fluid Steam, Air, H2, Fission Products Code Non-Code
- 8. ~EY Internal Volume, gal. 10 Design Pressure, psiq 100 Design Temperature, oF 200 Operational Pressure, psi~ 5 Operational Temperature, F 120 Material Stainless Steel 316L Fluid Post-Accident Reactor Coolant Code Non-Code
Table 9.3.11 (cont'd)
Desi n Data for Post-Accident Sampling S stem Com onents
- 9. Burette Internal Volume, cc 1000 Design Pressure, psig 100 Design Temperature, oF 200 Operational Pressure, psig 5 Operational Temperature, oF 120 Mater ial Stainless Steel 316L Fluid Post-Accident Reactor Cool ant Code Non-Code
- 10. Strainer Type "Y" Type Mesh Particle Size Retention" 250 Microns Operating Pressure, psig 2235 Operating Temperature, oF 621 Design Flow, gpm 2 Operating Flow (max) gpm 1 Clean aP (psig 8 gpm) 281 Loaded aP (psig 8 gpm) 1081 Collapse aP (psig 8 gpm) 7091 Gas Residence Chamber Design Pressure, psig 130 Design Temperature, F 350 Operational -Pressure, psig 80 Operational Temperature, oF i20 Volume,cc 4600 Fluid Post-Accident Reactor Coolant Material Stainless Steel 316L Code Non-Code
- 12. Exhaust Charcoal Filter Type Replaceable Cartridge Type Element Activated Charcoal Design Flow, scfm 333 Operational Flow, scfm 250-333 Operational Pressure Atmospheric Fl,ui d Aux. Bldg. Atmosphere Clean aP, inches water 9 scfm < 1 9 333 Loaded aP, inches water 8 scfm 1 8 333 Code Non-Code
. Table 9.3.12 Desi n Data for Post-Accident Sam lin S stem Process Instruments Instrument Descri tion ~Rccurac ~Ran e Boron Meter Density Sensor - 100 ppm 0 to 5000 ppm pH Meter Electrode Sensor - 0.05 3 to 12 Hydrogen Analyzer Thermal Conductivity - 2% of seal e 0 to 100%,
Sensor 0 to 10%
+-
Oxygen Analyzer Paramagneti.c Sensor 2% of scale 0 to 25%,
0 to 5%
I~
. Table 9.3.13 Instrument Calibration Fre uenc~
Component Calibration Maintenance Maintenance or Identification ~F ~F C lib ti T Charcoal Fil ter as req'd Replace filter when saturated, or when dosage is unacceptable (test with freon)
Pumps as req'd As required Valves 18 mos. Functionally test and repair as required Level Instruments 6 mos. Reset zero and span against known vessel levels Pressure Instruments 6 mos. Check accuracy against a standard Pressure Instruments 6 mos. Check pressure setooints with alarm 5 control functions pH Monitor 6 mos.* Calibrate with buffer solution H2 II o2 Meters 6 mos.-l yr Set zero and span using standard gases Boron Meter 6 mos.* Check zero, span, and temp.
compensator against test boron solution and de-mineralized water
~v Flow Meters 6 mos. Check accuracy against a standard Panalarm 6 mos. Check alarm function
- Calibration frequency can be extended until instrument malfunctions or gets unstable readings in a post-accident situation
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\%V ~ I ~ -~ ~ H 0 ~, =w. 'e i -v" h i 15.1.5 LINITIHG FAULT 3 EYENTS Limitin Offsite - of Nain Steam Outside Con-15.1.3.1
~ti",
~fT ii Il Dose Event Mii" Iii Loss ii i iiff it P ii i"
15.1.5.1.1 Identi ficati on of Event and Causes All Limiting Fault-3 event groups and event group combinations resulting in an increased heat removal by the secondary system shown in Table 15.1.5-1 were compared to find the event resulting in the maximum offsite doses. The loss of main steam-large, outside containment, upstream of NSIY with loss of off-site power as a result of turbine trip and with technical specification primary to secondary leakage ' through the steam generator tubes was identified as the limiting LF-3 event.
. ':-" The event groups and event group combinations evaluated and the signifi-cance of the offsite doses for each are'indicated in Table 15.1.5-1. All events indicated as insiginficant (I) would produce offsite doses well within the acceptance guideline in Table 15.0-4. All events indicated as significaht (S) produce offsite doses within the acceptance guideline.
The loss of main steam-large, outside containment may occur due to a break in the 34 inch main steam line.
Breaks ranging from 0.056 ft area up to the double-ended rupture of the 34 inch main steam line break are included in this event group. Events with break areas less than 0.056 ft2 are classified in the small loss of main steam event group. The offsite doses were maximized by assuming an interme-diate break (1.8 ft. ) which results in a minimum DHBR below 1.19. Technical specification tube'leakage also increased the offsite doses. The loss of of'fsite power as a result of turbine trip causes the coastdown of all reactor coolant pumps.
Of the two event groups, loss of main steam-large inside containment and loss of main steam-large outside corrfainment, in the LF-3 category, loss of main steam-large, inside containment will not cause a significant amount of steam release to the atmosphere and therefore will not result in significant off-site Poses. Loss of main steam-large, outside containment with a loss of offsite power and a technical specification tube leakage is the limiting event I
combination, since the decreased RCS flow due to the loss of power results P in degradation of fuel performance, and the technical specification tube leakage maximizes the release of activity to the atmosphere.
~
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15.1.5.1.2 Sequence of Events and Systems Operation Table 15.1.5.1-1 presents a chronological list and timing of system actions which occur following the large 1'oss of main steam event outside containmen with a loss of offsite power as a result of turbine trip.
The sequence of events and systems operation are indentical to those pre-sented in 15.1.5.3.2 and Figure 15.1.5.3-1 with the exception of the response of systems actuated by the occurrence of high containment pressure. High containment pressure is not present in this event.
Table 15.1.5.1-2 contains a matrix which describes the extent to which nor-mally operating plant systems are assumed to function during the transient.
The operation of these systems is consistent with the guidelines of Subsec-tion 15.0.2.3.
Table 15.1.5.1-3 contains a matrix which describes the extent to which are assumed to function during the transient.
safety'ystems
15.1.5.1.3 Analysis of Effects and Consequences a} Mathematical Models The NSSS response to a loss of main steam with loss of offsite power
~
as a result of turbine trip,was simulated using the CESEC computer program described in Subsection 15.0-4. The transient minimum DNBR values were calculated using the TORC code which used the CE-1 CHF, correlation described in Subsection 15.0-4.
. b) Input Parameters and Initial Conditions From the range of- values for each of the principal process variables given in Subsection 15.0-3, a set of initial conditions contained in 15.1.5.1-4 was chosen that produces the lowest minimum ONBR., 'able Additional clailification of the assumptions and parameters listed in Table 15.1.5.1-4,follows.
Maximum initial core power, maximum initial core inlet temperature, mimimum initial core mass flowrate and initial RCS pressure are chosen to minimize the ONBR, and maximize offsite doses.-
The moderator temperature coefficient and break size were varied to delay the occurrence of reactor trip either on low steam generator pressure or high core power level, thus maximizing the core heat flux.
An intermediate break size corresponding to 1.8 ft2 effective steam flow area per steam generator'ith a moderator coefficient of -1.6 x 10 -4 hP/F results in the lowest value of minimum DNBR and maximum degradation of fuel performance.
In order to further maximize the degradation in fuel performance and, thus, to maximize offsite do'ses, the time of turbine trip and the loss of offsite power, which caused four reactor coolant pumps to coastdown, is chosen so that the low reactor coolant flow trip condition occurs coincident r with the low steam generator pressure reactor trip.
In this event, the turbine is assumed to trip prior to reactor trip due to depressurization of Main Steam System. The reactor trip on low hdyraulic oil pressure is expected to occur during this event. In this analysis it 'is conservatNely assumed that this trip does not occur prior to reactor trip oaV .low -reactor coolant flow or low steam generator pressure.
The Pressurizer Pressure Control System and the Pressurizer Level Con-trol System are assumed to be in the manual mode of operation and, therefore, do not function to mitigate depressurization of the Reactor Coolant System {RCS). This results in low RCS pressure which mimimizes the DNBR.
The highest one pin radial peak with the most top peaked axial power stype is chosen to minimize the DNBR during the transient.
l' c) Results
~ The dynamic behavior of important NSSS parameters following this event are presented on Figures 15.1.5.1-2 to 15. Table 15.1.5.1-1 summarizes some of the important results of this event and the times at which minimum and maximum parameter values discussed below occur.
0 I A break in the main steam line outside containment causes an increase in steam flow, resulting in depressurization of the steam generators as shown on Figure 15.1.5.1-10. The pressure decrease initiates a low steam generator pressure trip and, subsequently, generates a maxn steam isolation signal (MSIS). HSIS closes the main steam isolation valves and main feedwater isolation valves isolating the intact steam generator while the steam generator connected to the ruptured line continues to blow down through the break.
The decreasing secondary pressure and temperature leads to an increase in primary to secondary heat transfer rate which causes the primary coolant (core average) temperature to decrease. Prior to reactivity due to a negative moderator temperature coefficient, the decreasing core average temperature causes moderator reactivity to increas, re-sulting in an increase of core power. After reactor trip, the core power further decreases to decay power level as shown on Figure 15.1.5.1-2.
The increasing core heat flux and the decreasing reactor coolant flow rate result in a decreasing minimum DNBR as shown on Figure 15.1.5.7-9.
The reactor trip causes the core heat flux to decrease resulting in a subsequent increase in minimum DNBR. The minimum DNBR experienced during a loss of main steam with a loss of offsite power as a result of turbine trip is 0.88/ resulting in 3.1 percent of the fuel pins in DNB.
During this event, two sources of radioactivity contribute to the off-site dose, the initial activity in the steam generator inventory, which is assumed to be 0.1 pCi/cc dose equivalent I-131, and the activity which 's added to the steam generator during the transient due to assumed Technical Specification. primary to secondary leakage through the steam generator tubes of 1 gallon/minute:
During the cooldown, steam. releases from the intact steam generator via the MSSVs and ADYs contribute to the offsite dose.
The offsite dose due to the loss of main steam-large, outside con-tainment with loss of offsite power and with technical specification primary to secondary leakage through the steam generator tubes results in no more than a 64 rem two hour inhalation thyroid dose at the ex-clusion area boundary. The total offsite doses during this event are shown in Table 15.1.5.1-5.
, 15.1.5.1.4, R
Conclusions This evaluation shows that the plant response to the loss of main steam-large, outside containment with loss of offsite power as a result of turbine trip and with technical specification primary to secondary leakage through the steam generator tubes results in maximum offsite doses which are within the acceptance guideline in Table 15.0-4.
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TABLE 15.1.5.1"1 SL2-FSAR SEQUENCE OF EVENTS, CORRESPONDING TIMES AND St&MARY OF RESULTS FOR A LARGE LOSS OF MAIN STEAM EVENT, OUTSIDE CONTAIRKNT UPSTREAM OF MSIV MITH A LOSS OF OFF-SITE P(NER AFTER TURBINE TRIP Success Paths 0
6 c
O 4J 44 0 0 O 0 O th 4J 4J M 4J 4J c(j 0 0 w'4 C$
v c 0 0 '0 Cl C4 cg 0 c bo gM 0 4J 4J 0 C CO 4J 4J Analysis U 4J 0 0 0 0 0 Q 4J 0 4J O g 4J Time Set Point CJ 0 C5 O 0 0 0 g a O 0 g i(4 C4 O c4 c4 40 H C4 H KK O H &w:
Sec Event or Value 0.0 1.8 ft2 break in a 34 'nch main steam line 47 Turbine trip assumed
- Off-site power lost
- Diesel generator starting signal
- Four RCPs coastdown Maximum reactor power, X 134 48.1 Reactor trip signal generated on low RCS flow, 'l. of rated flow or low steam generator pressure, psia 590
- 50. 9 Minimum DNBR 0.88 MSIS generated on low SG pres- 460 sure, psia 68.0 SIAS generated on low pre's- 1578 X X X X X surizer pressure, psia Pressurizer empties 130 HPSI flow begins 311 Affected steam generator empties 650 Operator actuates auxiliary feedwater to intact SG 15.1-li
TABLE 15.1.5.j.-l (Contcd) SL2-FSAR SEQUENCE OF EVENTS, CORRESPONDING TIMES AND
SUMMARY
OF RESULTS FOR A LARGE LOSS OF MAIN STEAM EVENTJ OUTSIDE CONTAINMENT UPSTREAM OF MSIV 1~1TH A LOSS OF OFF-SITE 2'OMER AFTER TURBINE TRIP Success Paths 0
EC 4J Cl 4J 4J CJ 0 CJ >
~Pl CJ CO 4J m 4J 4J C$ C CJ C 0 4J0 cU Q qj 0
CJ 4J g 00 4J 0 b0 4J 4J Analysis CJ e
44 4J CJ e
0 0 cp 4J 4J g $4 JJ CJ C 4J gM Time Set Point CJ V CJ CJ C 4 g 0 c C5 CC4 kl W O W Ww 44 Sec Event or Value CO 1800 1. Operator actuates atmos-pheric dump valves to commence cooldown of RCS
- 2. Operator loads the following on safety buses ~ ~
charging pumps pressurizer heaters
- 3. Operator borates to cold shutdown concentration
- 4. Operator clears SIAS and reestablishes letdown ,Lg 7200 Off.-site Power restored 12, 24(H. Shutdown cooling initiated, 350/275 X F/psia C ~
15.1-
TABLE 15.1.53.-2 SL2-FSAR DISPOSITION OF NOR'.lALLY OPERATING SYSTEi~lS FOR THE LOSS OF HAIN STER~i-LARGE OUTSIDE CONTAIhÃiENT UPSTRE OF MSIV WITH THE LOSS OF OFFSITE POWER AFTER .TURBINE RIP- ~~
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- 1. Main Feedwater System X
- 2. Turbine-Generator Control S stem
- 3. Steam Bypass Control S stem X
- 4. Pressurizer Pressure Control System X
- 5. Pressurizer Level Control System
- 6. Control Element Drive Mechanism Control System
- 7. Reactor Pegulating System
- 8. Reactor Coolant Pumps
- 9. Chemical and Volume Control S stem
- 10. Condenser Evacuation System
- 12. Component Cooling Water System X
- 13. Turbine Cooling Water System X
- 14. Intake Cooling Water S stem X 2 4
- 15. Condensate Transfer System
- 16. Circulating Water System
- 17. Spent Fuel Pool Cooling System~
- 18. AC Power (Hon-Safety) X
- 19. AC Power (Safety)
- 20. D. C. Power X
- 21. Power Operated Relief Valves X
- 22. Instrument Air S stem
- 23. Haste Management-Li uid System has no automatic mode.
Lose power on loss of offsite power, then automatically loaded on diesel generator.
- 3. Operator must connect to safety bus for operation.
- 4. Only essential portions of the system are available.
15.1-.
TABLE 15.1.5.1-3 SL2-FSAR UTILIZATION OF SAFETY SYSTB"8 FOR THE LOSS OF MAIN STEAM-LARGE. OUTSIDE CONTAIhPi'KNT UPSTREAM MSIV LOSS OF OFF SITE PO>~ER AFTER TURBINE TRIP PY C7 gx g + M I
- 1. Reactor Protection S stem
- 2. En ineered Safet Features Actuation S stems
- 3. Diesel Generators and Su ort S stems
- 4. Reactor Tri Switch Gear
- 5. Main Steam Safet Valves
- 6. Pressurizer Safet Valves
- 8. Main Feedwater Isolation Valves
- 9. Auxilia Feedwater S stem X
- 10. Safety In ection S stem ll. Shutdown Coolin S stem CCW & ICh'2.
Atmos heric Dum Valve S stem
- 13. Containment Isolation S stem
- 14. Containment S ra S stem
- 15. Iodine Removal S stem
- 16. Containment Combustible Gas.Control S stem
- 17. Containment Coolin S stem' NOTES:
- Manually actuated during normal cool down
- 1. Normally operating system (in nonsafety mode) 2~ Permissive blocks of SIAS and MSIS are manually actuated to permit shutdown depressurization.
Systems not checked are not utilized during this event.
15.1-
SL2-FSAR TABLE 15 ' ~ 541-4 ASSI%i ED INPUT PARAMETERS AND INITIAL CONDITIONS FOR LOSS OF NAIH STEAN~
A RESULT OF TURBINE TRIP Parameter Assumed Value Initial Power Level, MWt 2621.4 Initial Core Inlet Coolant Temperature, F 551 Initial Core RCS Flow Rate, gpm 370,000 Initial RCS Pressure, psia 2,150 Initial Pressurizer Water Volume, X Level 53 Axial Shape Index -0 '
Doppler Coefficient Multiplier lio Moderator Temperature Coefficient, 10 bp/F CEA Worth for Trip, 10 hp Breal; Size, ft 15.1>>
SL2-FSAR TABLE 15.1.5.1-5 OFFSITE DOSES Two Hour Exclusion Area Entire Event Lou Boundary Dose Population Zone Dose Thyroid 64'em Whole Body 15.1-
150 120 90 K
o~ 60 0
0 360 720 1080 18CC TIME, SECONDS FLORIDA POWER 8 LIGHT COMPANY ST. LUCIE PLANT UNIT 2 CORE POWER VS TIME h ~
FlGURE 15.].5.l 2
4 ir I 150 D
FJO 0
U I-lLJ x 90 U
KO LIJ K
60 0
O 8 I-30 0 360 =
720 1080 1440 1800 TIME, SECONDS y ~
, h et FLORIDA POWER K LIGHT COMPAHY ST. LUCIE PLAHT VHIT 2 CORE AYERAGE HEAT FLUX VS TIME FIGURE 15.1.5,l 3
3000
<<DOES NOT INCLUDE ELEVATION AND REACTOR COOLANT PUMP HEAD EFFECTS
~:2400 1800 I-Z cK o "1200 O
I-O 60Q 0 720 1080 i440 'BGC TIME, SECONDS FLORIDA POWER 8 LIGHT COMPANY ST. LUCIE PLANT UNIT 2 REACTOR COOLANT SYSTEM PRESSURE YS TIME FIGURE .15.1.5. I-4
A 4
c 7
U O OUTLET CO SOO p
Lll o AVERAGE 400 INLET 0
0
~o '0C 200 360 ~ 720 10SO . 1440 1SCO TIME, SECONDS FLORIDA POWER E. LIGHT COMPANY ST. LuCIE PLAHT UHIT 2 CORE COOLAHT TEMPS. YS, TIME FIGURE 15.1.5.I-5
k' MODERATOR DOPPLER SAFETY INJECTION TOTAL CEA 360 720 i440 1800 SECONDS
'IME, FLORIDA POWER E. LIGHT COMPANY ST. LUCIE PLAHT UHIT 2 REACTIVITY YS TIME t=lGURE 15.1.5.f-6
~ '
800 600 400 200 0
0 360 . 720 1080 1440 1800 TIME, SECONDS FLORIDA POWER L LIGHT COMPANY ST. LUCIE PLAHT UHIT 2 PRESSURIZER WATER YOLUME YS. TIME FIGURE 15.1.5.I-T
1 ~ 2 1 ~ 0 0 0
~
I 0 360 .720 1CBO .'40 TIME, SECONDS
'p FLORIDA POWER 8 LIGHT COMPANY ST. LUCIE PLAHT JHIT 2 REACTOR COOLANT FLOW YS TltAE
1.8
~ '
zO S.2 R
0.9 0.6 0 10 20 30 50 l IME, SECONDS FLORIDA POWER 8 LIGHT COMPANY ST. LuCIE PLAHT UHIT 2 MINIMUMDHBR YS TIME FIGURE 15.1.5./-9
1000 800 INTACT LINE 400 RUPTURED LINE 200
~ r 0 360 720 1080 1440 1800
~
TIME, SECONDS lI, FLORIDA POWER L I IGHT COh(PAHY ST. LUCIE PLAHT UHIT 2 STEAM GEt<ERATOR PRESSURE VS Tlh(E FIGURE 15.1.5. I-10
I 300 246 CO K
a 180 120.
Ql INTACT LINE 60 RUPTURED LINE 0
0 720 1080 1440 1800 TIME, SECONDS FLORIDA POWER It LIGHT COMPANY ST. LUCIE PLANT OHIT 2 STEAM GENE RATOR LIQVID MASS YS TIME FIGURE 15.1.5.I-11
. 0 4000 2400 1600 800 RUPTURED L)NE INTACT L)NE 0 360" 720 1080 1440 1800 TIME, SECONDS
- ~.
o> CO 3c 0 w I-
. ~~5 2ap LX g u) LQ~
I-
~
160'0 0 360 ., 720 1080 i 440 'BGQ TIME SECONDS c ~
FLORIDA POWER L LIGHT COMPAHY ST. LUCIE I'LAHT UHIT 2 IHT EGRAT ED STEAM FLOW YS TIME FIGURE 15.1.5. I-13
2000 1600 120C 800 400 0 360 -. 720 1080 1440 180C TIME, SECONDS FLORIDA POKIER It LIGHT COMPANY ST. LUCIE PLANT UHIT 2 FEE DNA TER FLOW YS TIME FlGURE 15.1.5, l-14
A, 500 400 30C x
200 100 0
0 360 '20 1080 .1440 , 8r.n TIME, SECONDS FLORIDA POWER Il LIGHT COMPANY ST. LuCIE PLANT UNIT 2 FEEDWATER ENTHALPY VS TIME
. FIGURE lS.1.5.l-lS
H. IVugiN RENts~cQ <0 FKLi5 AQa~g se.5 I 0 Res>>4~K SL2- FSAR g~U ~- og g7-"g FLo~ IZED.QU i RG 8 E QT< ~
between core heat addition and steam generator heat removal prior to the CEA insertion and, hence'o maximize the peak RCS pressure. The
.'affected steam generator is assumed to instantaneously lose all heat transfer capacity when total depletion of its liquid inventory by boil-off and discharge occurs. The break area, which resulted in
- the highest peak RGS pressure, was found to be 0.25 ft ~
'Using the highe'st initial core power maximizes the RCS heat"up which is the driving force of the pressurization. Variations of initial core inlet temperature and initial- reactor coolant flow had negligible effects on the peak RCS pressure. The highest initial core inlet temperature and the lowest initial reactor coolant flow were used xn the analysis. The Pressurizer Pressure Control System is placed in the automati.c mode, such that it delays reactor trip, thus prolonging the RCS heat-up and increasing RCS pressurization. Using the smallest CEA worth ana the least negative moderator temperature coefficient.
maximizes the heat flux overshoot after reactor trip, increasing the
~
RCS heat-up.
The highest initial pressurizer liquid volume and manual operation of Pressurizer Level Control System were used to allow the maximum in-crease of. pressurizer level, maximizing the transient effect of RCS pressure increase during heat-up. However, the selection of Press-urizer Level Control System operating mode and initial'ressurizer liquid volume has only a small impact on the peak RCS pressure.
Auxiliary feedwater was assumed to be activated by the plant operator within f've'minutes of the low steam-generator level trip condition to prevent the pressurizer from filling solid. The assumed flow to the intact steam generator is 500 gpm.
( The peak 'RCS pressure occurs at 31.6 seconds. Analysis has shown that one motor-driven AFll pump automatically starts deIivering if only 320 gpm to the intact steam generator at 146 seconds, the peak pressure will be unchanged hnd the pressurizer will be prevented from filling solid.)
To maximize RCS pressure, the is modes SBCS assumed to be in the manual In oraer to eliminate the impact of uncertainty. in the water C
the affected steam generator, reactor trip on a low level of assumed to occur until dryout of the affected water level is not steam generator.
It is anticipated that equipment may be actuated by high containment pressure during this event. These actions are identified in the sequence of events, but are conservatively assumed quantitative analysis of the NSSS response to this event. not to occur in the
Q c) Results The dynamic behavior of important NSSS parameters following loss of feedwater inventory with loss of offsite power as a result of turbine trip is presented in Figures,)5.2.5.2-.2 to 20.. Table )5.2.5.2-) .
summarizes some of the important results of this event and the times at which the minimum and maximum parameter values discussed'elow occur.
A rupture in the main feedwater line instantaneously terminates feed-water flow to both steam generators and causes liquid flow from the
I 4
Pe have ra~ lcd Subsection 10.2.3 of the Final Safety Analysis R port submitted by the applicant. Our evaluation cannot be completed vi "hout additional ink ometion frown the applicant reiating to the design, assembly and operating conditions of the lov. pr" ssu-c turbine. discs. Past experience pith similar equipment in the &iced Kingdom and more recently vith Vestinghouse turbines in the United States has revealed a propensity for stress corrosion cracking in discs which @as not predictable. Xn order for the staff to assess the potential for stress cor"osion cracking in the applicant's plant, the following inxormatio"., vill be requixedi a) baat lubricant vas used in the hub area of the discs for a ss cub ly.
b) Mhat are the similarities jdi fercnces between the discs in the St. Lucie Bait Ho. 2 turbines aaa those used by VQst inghous e+
Feat are the operating, temperatures in the bore area of the discs.
%hi& disc or discs are exposed to a moisture level during operaticn that sppxeximates the level of moisture present in cases of cracking~
l'hat ar the calculated critical crack sizes and what is the method used, to calculate that sisc.
>~hat c~abiLity for volumetric inspection of the disc hub areas "-'s ave lable to Sc ~ Xacic Unit 5o. 2.
The Dxbricant used in the hub area oX the discs for sssembl>. is Wlybdenum Disulfide or a Graphite Mbricant.
St~ ~mcxe Unxt Boi 2 LiP. turbine discs are sxmx.lar in design to other Westinghouse units. Discs are'hrunk on and keyed by means of three keys to the shaft- .
c) 'ihe operating temperatures and moisture level for the
.various discs in the bore area are as folio+ax P
252'-I Amend.-.ent No. 6, t;9/Sl)
Tllle t Bore Rx1 t Dora kL~ ta 1 Tcpp F Whtal T{'np F Inlet Stcam (hale t (Tate 1) {i<~{>> 2) )Sig>> s ter Steam ~u9.s-"uW 926 0 1.8 2 280 263 1.8 6.2 3 228 209 6~2 8 5 4 191 181 8>>5 9>>8 5 186 186 9+8 Xl.9 khte (1) 2" from Xnlat disc face
Ãote (2) Hid key location o". disc outlet edge V. %e table above Eor r;oisture level or" Various discs du"i'peration.
E. Cri,t;ical crack size is calcul.".Red to be as follows.'.P.
Cx1t1cal CEQCJ{, 83.58 Crxtacal Crack 81"e No>> Di>>>> Ra. Bore {in i>>ch>>s> ia ia>>a (ia i.acbes)
L>> P. /e. 1 Disc !5 l79 2>>665 Go'acrnar cad L.P. No. Disc 2 Govcloor and 2.696 l. 207 L. P>> Disc 3 4. 243 , 2.116 Govezoor cAd L>> P. tb>> Disc 4 4.961 2.5>V Covexaor e<>d L>>P>> le. 1 Disc 5 3.973 1.957 GQveX'rior Clld X.P>> bo. 1 Bise 1 2.573 l>>1M GGQ>> 6'Qd L. P, Ho. 1 Disc 2 2>> 220 o,e26 Cea. end LP>> R>> 1 Disc 3 3.534
. GBQ>> akid 1.69'-839 L>> P>> 5o. 1 Disc 4 9-Q>> Gad J.P. Fh 1 Qiac 5 8 204 4>>440 C~Q>> CQQ L>> P. tk)>> 2 Disc 1 2.709 1>> 215 Coveraor end X>>P bo. 2 Disc 2 2.196 0>>914 Caveroor end
C'critical Crack Sire Crx txcQl Cxsck>> Lee L. P ?io. DL sc No>> Lore (in inc?ll ~) Lil KP>>43" (XTI Lneiles)
L.P i fiO>> 2 Dxsc 3 3.69/ X. 795 CCvtÃPQx'H8 X.P. ?h. 2 9'ee 4 6.483 3.430 Clove%'Qcz Clio I:P No. 2 Disc 5 7.8l3 4.211 GOVT'X'J30l Q338 L.P. Viol 2 DLSC 1 2.66C 1 l86.
Cca e 338 I:P. h0. 2 Disc 2 2. 233. 0.936 68n>> end L P>> ho>> 2 Disc 3 3.6~<8 l,766 C~n. end it ~ P ~ I'4>> 2 Disc Il 4.749 2.412 Gcn ~ Ggd L.P. R ~ 2 Disc 5 8.237 4.460 CC~< ~ QACi TAe clet"Iod 0 c81cuiet. cN3's eel'. ii e.? i 'l Peed-3 33gJlouse Rel301 t
?'vcleax 7orbine ksc 33~~pec,'ioll" sui~";i::tel3 .0:l?'C in Jane, 1981, Discs X sad CZLtetxe Rot L P 2
epprcxiznat.es the lave 0: -,.Cis;'u'.e o:" f."Iit. xn cclses cf ctacicing ~
F..he BorRB elle Rf vl"BYE CAil be i33'..PCCt::l:-,, uitraS033i C illSPeCti033 teehiliques l>>it!Iollt x'eeovinp:he disc Srcr3 rt e s-.i~r't. St- Llicie Unit ?io ~ 2 his bBQA subjected ko this inspection at the sl.te.
SER ITEN 3 Question No.
420.05 The instrumentation and control system comparison information of (7.1) FSAR Table 1.3-1 and FSAR Subsection 7.1.1.6 is insufficient.
The information supplied does not completely show that each instrumentation, control and supporting system is:
- 1. Identical to that of a nuclear power plant of similar design which has recently received an operating license, or
- 2. Different from previous/recent designs with a discussion of the differences and their effects on safety related systems.
The above information is required by Regulatory Guide 1.70.
Revision 3, "Standard format and content of safety reports for nuclear power plants", Section 7.1.1. Therefore, in conformance with Regulatory Guide 1.70, Section 7.1.1, provides a comparative discussion for each St. Lucie 2 instrumentation and control system.
~Res onse Chapter 7 will be revised to include the following information:
St. Lucie 2 instrumentation and control systems are designed and built functionally identical to those systems provided for St.
Lucie 1 (Docket Ho. 50-353). In addit'on section 7.1.1.6'ill be amended as follows. Also see table 420.05-. 1 for detailed compari-
.sons of the Reactor Protection System.
SL2-FSAR Shield Building Ventilation System Switchover from Fuel Handling Building (E) (see Subsection 6.2.3)
The above are described further in'Section 7.6.
7.1,1.6 Comparison'he Reactor Protective System was designed and built functionally identical to the system provided for St Lucie Unit 1 (Docket No, 50-335) with the following exceptions:
The number of CEAs is changed to 91. The corresponding change in the number of CEAs and CEA subgroups has resulted in minor changes in CEA arid
. KR53 b) . RPS St. Lucie Unit 2 has a loss of CCW trip for RCP (Equipment) protec-tion. This trip is neither presently nor previously licensed, and is not credited in the safety analysis.
Overall the Nuclear Instrumentation portion of the RPS is function-ally identical to the St. Lucie Unit 1. However, the sub-function, Zero Power Node Bypass (ZPIB) is initiated by the low power section of the Linear Power Range Safety Channel rather than the Wide Range Log Channels, as in St. Lucie Unit 1. The St. Lucie Unit 2 Wide Range Log Channel will have an accuracy of + 3% of full scale voltage (corresponding to a factor of two in absolute power level); with a response time of 50 msec. The St. Lucie Unit 2 design specifies initia-tion of ZPi% by the low power circuitry of the Power Range Safety Channel.
Here, a smaller linear range is utilized to provide improved setpoint accuracy and better response time. In particular, the low power circui-try has an accuracy of + 0.1% of full scale power over its range of 102 to 2% power; with a response time of less than 25 msec.
The method of ZPMB initiation employed in St. Lucie Unit 2 is superior to that used in St. Lucie'nit 1 since the setpoint accuracy and the response. time have been improved by a least a factor of two.
4 The that used for St. Lucie Unit St. Lucie Unit 2 Logic functions are identical toconnections as part of 1, but also includes fuses in all matrix a inter-baytest circuit is provided for improved fault protection. In addition, periodically( (2) checking the fuses 4ssociated with this matrix fault accordance with the protectiorij Matrix fuse integrity will be checked periodically in RPS technical specifications.
dry reed types, for improved re-St. Lucie Unit 2 matrix relays are wetted reed type liability over the original St. Lucie Unit 1 mercury r clay design. bistable design w St. Lucie Unit 2 incorporates a new RPS by: greater accuracy, input identical, is characterized improved'noise immunity improved circuit isolation,
'unctionally b u ffering f or h-vi an adjustable response time, (down)
'a less cvcling due to a va riable forces i ich sterisis feature and a pull-up signal.circuit design Uh a Consequently, contrary trar to the bistable trip on a loss of input Si'. Lucie Unit 2 auction-St. Lucie Unit I FSAR Section 7.2.2.2, theinputs will trip in an open eered Input bistables ~tilizing negative circuit configuration.
E St. Lucie Unit 2 has incorporated RG1.53.RG1.22, RG1.75, IEEE-323-74
-344-g5 and -384 74 in the RPS design, as these standards were not in effect when St. Lucie Unit 1 was licensed.
c) ~Sstems Re uired for Safe Shutdown St. Lucie Unit. 2 conForms to RG 1.75, which identifies a 6-inch spatial separation requirement, versus the 12 inch criteria of St. Lucie Unit l.
/
d) Safet Related Dis la Instrumentation
- The upper and lower CEA limits are indicated on the CEDMCS panel for St. Lucie Unit 2, while St. Lucie,Unit 1 displays this information on 'the core mimic display. The S" .'ucie Unit 2 design is identical to the SONGS design (Docket No. 50-362).
- Many aspects-of the St. Lucie Unit 2 design for Post Accident Monitoring .
are different from St. Lucie Unit 1. St. Lucie Unit 2 is identical to SONGS with the exception of. invoking BTP EICSB No. 23, gualification of Safety Related Display Instrumentation for Post Accident and Safe Shutdown. The associated changes in this area for invoking RG1.97, Rev 02 are provided as part of response to ICSB question 420.41.
St. Lucie Unit 2 utilizes the Analog Display System (ADS), which while functionally identical to the St. Lucie Unit 1 Metroscope, exhibits improved reliability design features and incorporates improved human factors characteristics.
This section is not 1E indication and is further described in Section 7.7.1.1.6.
e) En ineered Safet 'Features Actuation S stem'(ESFAS)
The St. Lucie Unit 2 ESFAS is functionally identical to the St. Lucie Unit 1 System. Channel designation and parameter inputs are essentially the same except for the following specific differences. The St. Lucie Unit 2 main steam isolation signal (MSIS) is initiated by a low pressure signal from either steam generater or high containment pressure (FSAR section 7.3.1.1.5).
St. Lucie Unit 1 MSIS is initiated by a low pressure signal from either steam
generator only. .The St. Lucie Unit 2 containment isolation actuation signal (CIAS) wil1 be modified to actuate on'safety injection actuation signal (SIAS) as well as high containment pressure or high containment radiation.
The modification was incor+rated in St. Lucie Unit 1 as required by USNRC
- 'HI Action Items to satisfy a diversity requirement for containment isolation.
FSAR section 7.3.1.1.4 will be revised to reflect this CIAS modification.
St. Luqie Un'it 2 has incorporated R.G. 1.53, 1.22, 1.75, IEEE 323-1974, 344-1975, and 384-1974 in the ESFAS design, as these 'st'andards were not in effect when St. Lucie Unit 1 was licensed.
NOTES
- 1. This design reflects that of the C-. E. System 80 design, (Docket No. STN-50-470F),
wh'ich is qualified to IEEE 323/74, and 344/75.
- 2. The fuses of this section are utilized in the System 80 Plant Protection System Design.
- 3. The bistable design is a modified System 80 design, since the System 80 design does not utilized auctioneering.
I
The ESF systems was designed and built functionally identical to the ESF systems used on St Lucie Unit 1 (Docket No. 50-335). The following are ESF system differences when compared against St Lucie Unit l.
a)'t Containment fan cooling system has two speed motors.
I b) HPSI and LPSI pumps are provided with redundant mini flow (re-circulation) headers.
c) LPSI 'pumps have redundant headers and associated valves.
d) The HPSI pumps are comprised of two functionally separate and independent pumps and headers.
e) The Shutdown Cooling System is designed with redundant valves and headers.
Piping and valves permit the diversion of HPSI flow from the cold leg into the hot leg of the Reactor Coolant System.
7.1.2 IDENTIFICATION OF SAFETY CRITERIA Comparison of the design wi tb'applicable Regulatory Guide recommendations and degrees of compliance with the appropriate design bases, criteria standards, and other documents used in the design of the systems listed in Subsection 7.1.1 are described in Subsections 7.1.2,1 through 7.1,2,2.
- 7. 1.2. 1 Desi. n Bases The technical design bases for specific instrumentation and controls of each safety related system are presented in applicable subsections of this chapter, Design bases that apply equally tq all safety related instrumentation and control systems are in this subsection.
Table 420.05-1 Page 1 of 8 St. Lucie I and II RPS Comparison St. Lucie I & II Comparison (Standards, Functional & Hardware)
& St. Lucie II New FSAR Section Additions Comment Other Licensing Applications 7.1 General functional differences are Reg Guides and Standards are identified for the RPS and See Section 7.1.1.6 similar to Songs ESF in Section 7.1.1.6.
Except for those identified, St. Lucie II is functionally
. identical to St. Lucie I.
Some safety criteria differ-ences are not apparent because of the St. Lucie I and II FSAR format differences. However, St. Lucie II is explicit 'in identifying by comparison the Regul'atory Guides, versus the St Lucie II INC design basis.
0 Table 420.05-l Page 2 of 8 St. Lucie I and II RPS Comparison St. Lucie I II Comparison (Standards, Functional & liardware)'
Section St. Lucie II New Commi nt Other Lzcenszng Applications FSAR Additions St. Lucie II does not incorpor- St Lucie II is not a Not required for 2560 HWT 7.2 ate Asymetric Steam Generator "stretch" power plant operation.
Tilt (ASGT) as part of the TH/L therefore ASGT is not a PVAR calculation design requirement 7.2.1.1.1.4'.2.1.1.1.6.
St. Luci e I I has provisions for St. Lucie I 'resently installed in i.
3 E 4 RCP operation.
RCS flow trip setpoint is The low Same as
't. Lucie I switched from the 3 pump to the 4 pump setting. automatically when reactor power is >605 Loss of CCl< trip is added to Desired by FP8L II Not previously or presently 7.2.1.1.1.11: St. Lucie II for RCP (equipment) licensed;: Not credited in al so protection J the Safety Analysis.
7.2.2.2.10 Coranents included in Tables 7.2.1.1.2.1 7.2-3 and 7.2-4 Overall the Nuclear Instr.
7.2.1.1.2.2 System performs functionally identical to St. Lucie"I.
Certain sub-functions are performed differently in St. Lucie II. These are as follows:
(1) The Zero Power Yiode Bypass This eliminates the need Not previously licensed.
is initiated by the low to qualify the.Mide Range Reflects System 80 design power section of the Log Power Safety Channel. 'nd qualification linear Power Range Safety Channel The Excore Safety Channels are New qualification standards 7.2 Per System 80 Design to IEEE 323-74 and,, 'ualified IEEE 344-75
Table 420.05-1 Page 3 of 8 St. Lucie I and II RPS Com arison St. Lucie I & II Comparison (Standards, Functional & Hardware)
& St. Lucie II New FSAR Section Additions Comment Other Licensing Applications T.2.1.1.3
- l. St. Lucie II Logic functions For fault protection via Same .design as System 80-PPS.
i denti cal to St. Luci e I. inter cabinet wiring Not previously licensed.
St. Lucie II however has fuse in all matrix inter bay connections.
- 2. St. Lucie II Matrix relays are dry reed types. St. Lucie I High failure rate of Dry reed relays have 'been mercury-wetted types in supplied to Maine Yankee are mercury-wetted types. older plants prompted Omaha, Millstone II, and replacement with dry reed. possibly St..Lucie I for replacements.
7.2.1.1. 4 The reactor trip switch gear. llhile functionally The RTSG is identical to no ldnger houses the under'- the relays are now identi-'cal, Songs. In addition, IEEE 384 voltage relays for auxiliary . housed in the CEDMCS has been incorporated for functions. cabinet St. Lucie II.
7.2.1.1.9.9 St. Lucie II has additional Yerify continuity of all Not licensed on any other test circuitry for checking th fuses periodically. plant nor in System 80 design.
fuses of the matrix fault St. Lucie I does not have protection fuses.. fuses to test. Test circuit only. Open fuse is conservative.
~
7.2.1.1.9.2 Trip test cable selects (plugs This section is identical This was explained to clarify into) measurement channel. The to St. Lucie I. the St. Lucie II design as B/S auctioneer select switch is used to select the auctioneered worded in the FSAR.
channel to auctioneering input '
bistables only.
Table 420.05-1 Page 4 of 8 St. Lucie I and II RPS Com arison St. Lucie I & II Comparison (Standards, Functional & Hardware)
& St. Lucie II New FSAR Section Additions Comment Other Licensing Applications
- 7. 2 Cont' St. Lucie II incorporates a St. Lucie II has a NN licensed previqusly.
OTHER new bistable design which has'. different manufacturer for these bistables. Thus the This is a modified System 80 (1) greater accuracy benefits of 1-4. design, incorporating auctioneering, since System 80 (2) circuit isolation via input does not utilize auctioneering.
buffers (3) adjustable time response for noise immunity (4) Variable hysteresis to
'. prevent cycling (5) Pull-up (down) circuitry (5) St. Lucie I FSAR
~ which forces a bistable '. indicates that auction-trip on a loss of input eered input bistables signal, i.e. open circuitry uti.lizing negative will not trip'.inputs These units are directly in an open circuit interchangeable with the old -.!. configuration design and are functionally (St. Lucie I FSAR identical 7.2.2.2)
R.G. 1.53, Single Fai.lure Crit. These documents were not in 7.2.1.2 R.G. 1.22, Periodic testing effect for St. Lucie I.when IEEE 323, 344 licensed.
IEEE 279/71, 384/74, and IEEE 384 and RG 1.75 were 7.2.2.3.2 RG 1.75/75Rl statements of not in effect when'..
conformance of design St. Lucie I was licensed.
IEEE 279/71 was in effect but conformance is more explicit in the St. Lucie II FSAR
C Table 420.05-1 Page 5 of 8 St. Lucie I and II RPS Com arison St. Lucie I & II Comparison (Standards, Functional & Hardware)
& St. Lucie II New FSAR Section Additions Commi nt Other Licensing Applications Tabl e Components are identified for FPSL design addition Not:credited in the Safety 702 "3 RCP and CEDMCS CCH Analysis Equipment protection only.
.'able (1) St. Lucie II does not This is the same format
,7. 2-4 accuracy and list'hannel Songs'SAR has.
i response times.
(2) RCP and CEDMCS CCW channel addition. Not credited in Safety
'.is added here FpgL design Analysis..'quipment Protection only.
(3) Listed parameter ranges are Later design, closer to System 80 slightly different than St. Lucie I'. Axial shape index is omitted.
System 80.
Table 420.05-1 Page 6 of 8 St. Lucie I and II RPS Comgarison St. Lucie I & II Comparison (Standards, Functional & Hardware)
& St. Lucie II New FSAR Section Additions Commi nt Other Licensing Applications 7.4.2.2 St. Lucie II identifies a 6 Lucie'I conforms to System 80 inch spatia1 separation between RG 1 75, which identsf>es redundant circuitry while the 6 inch separation St. Lucie I identifies a requirement.
12 inch separation.
7.4.2.1.2 St. Lucie II refers to This cr iteria is RG 1.75 (in FSAR for single failure criteria power't.
Section 8.3) from St. Lucie I slightly'ifferent criteria since RG 1.75 was associated with 1E channels not in effect when during a loss of offsite St. Lucie I was licensed.
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Table 420.05-1 Page 7 of 8 St. Lucie I and,'II RPS Comoarison St. Lucie I 'lI 5
Comparison (Standards, Functional & Hardware) 6 St. Lucie II New FSAR Section Additions Comment Other Licensing Applications
~
7.5
'7.5.1. 4c St. Lucie II upper and lower The SONGS design is Songs-CEA limits (reed switch) are implemented for St; Lucie functionally identical to 1
II.
St. Lucie I. Indications of these limits are .located on the St. Lucie II CEDMCS control panel versus the Core Mimic as in St. Lucie I.
Post Accident Monitoring for St; Lucie II is identical Songs. =-
7.5.1.8 St. Lucie II is different in to Songs with exception of However, RG 1.97 will be and many aspects when compared to invoking Branch Technical invoked in these areas for 7.5.2.9 S.t. L'ucie I Position EICSB No. 23, St. Lucie II. The associated gualification of Safety changes are forthcoming.
Related Display Inst. for Post Accident and Safe Shutdown.
St. Lucie II incorpor'ates The Analog Display System Not a licensing requirement.
7.5.1.4 the Analog Display System, has improved reliability which is functionally identical design features as well to the- St. Lucie I Metrascope. as improved human factors characteristics.'
Table 420.05-1 Page 8 of 8 St. Lucie I and II RPS Coa: arison St. Lucie I & II Comparison (Standards, Functional & Hardware)
& St. 3 ucie II New FSAR Section Additions Comment Other Licensing Applications 7.7.1.1.1 CED)PCS is functionally St. Lucie II is a 16 by 16 Songs identical except as noted in . core with 91 CEA's. The Section 7.7;1.2.1. =
differences identified in 7.7.1.2.1 inhere implemented to incorporate Human Factors Engineering.
7.7.1.1.8 The Incore Instrumentation St. Lucie II is a ANO 2 System is similar to that ) 6 x 16 core.
supplied for AN02, except that 56 detectors are utilized on St. Lucie II versus 44 on ANO 2 (Section 7.7. 1.2.8)
St'. Lucie II has 4 linear 7.7.1.2.9 of System 80 7.7.1.1. 9 See power range/log wide range Lucy II FSAR See also 7 7 1 2 9 of safety channel drawers in St. Lucie II FSAR; ;
In addition they have the.'PS.
2 startup/control channels for non-safety operator func.ions for start up and steady state power operation load following. 'nd The St. Lucie II source range (startup) channels use BF3 detectors instead of the B10 used on St. Lucie I.
8.3 Item 90%
Additional Information to Existing Response on SER Open
. 480V Motors.'he load sequencing (timing) relays used on St. Lucie Unit 2 are the same as those used on St. Lucie Unit 1. Limited set point drift experienced with St. Lucie Unit 1 timing relays has resulted in an ongoing evaluation of solid state electronic timing relays having greatly improved repeat accuracy as replacements for the existing relays. Experience on St. Lucie Unit 1 has shown that set point drift does not exceed +1 second in 18 months and that this drift has never impacted the starting and loading of the diesel generators. The preclude any possibility of load sequencing relays will be verified operable with the interval between each load block within +1 second if its design interval every 6 months. Upon satisfactory completion of the electronic relay evaluation, the existing relays will be replaced during the first refueling outage. Replacement of the existing relays with low drift electronic relays shall be cause to return the survel-lence requirement to once every 18 months during plant shut down.
Chapter 8.3 SER Item . Isolation Devices The St. Lucie Unit 2 design will be modified such that those non-safety loads connected to the Class 1E busses which are not considered extremely impor-tant for operation and plant investment will be shed from the Class lE busses by a Safety Injection Signal or will be locked out of service during plant operation in accordance with the Technical Specifications.
Those non-safety loads which are considered extremely important for operation and plant investment will remain connnected to the Class lE busses, however, they will be provided with two, Class lE, Fault current interrupting devices.
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