ML18038A244
| ML18038A244 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/18/1987 |
| From: | Mangan C NIAGARA MOHAWK POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-C-08, REF-GTECI-OS, RTR-NUREG-1169, RTR-REGGD-01.096, RTR-REGGD-1.096, TASK-C-08, TASK-C-8, TASK-OR (NMP2L-1007), NUDOCS 8703200296 | |
| Download: ML18038A244 (149) | |
Text
REBULAT INFORNATION DIBTRIBUTIOtYBTEN'RIDE)
ACCESS~>QN NBR: 8703200296 DOC. DATE: 87/03/18 NOTARIZED: NO DOCKET O FACIL: 50-410 Nine Nile Point Nuclear Station) Unit,2s Niagara Noha 05000410 AUTH. NANE AUTHOR AFFILIATION NANQAN, C. V. Niagara Nohaek Poi er Corp.
RECIP. NANE RECIPIENT AFFlLlATION Document Control Branch (Document Central Desk)
SUBJECT:
Requests that, NRC accept alternate to Reg Guide i. 96>
"Design of Nain Steam Isolation Valve Leakage Control Bijs for Boiling Water Nuclear Pokier Plants" as. described in Attachment A> per 870318 discussion. Related info enc l.
DISTRIBUTION CODE: A003D COPIES RECEIVED: LTR ENCL SIZE:
TlTLE: QR/Licensing Submittal: Supp l 1 to NUREQ-0737(Qeneric LtT 82-33)
NOTES:
RECIPIENT COP IEB RECIP I ENT COP IEB ID CODE/NAl'lE LTTR ENCL ID CQDE/MANE LTTR ENCL BWR ADTS 1 BWR EB 1 1 BWR EICBB 2 2 BWR FQB 1 BWR PD3 LA 1 BWR PD3 PD 7 7 HAUQHEY. N 1 BWR PSB 1 BWR RBB 1 INTERNAL: ADN/LFNB 1 0 IE/DEPER/EPB 3 3 NRR BWR ADTS 1 1 NRR PWR-B ADTS 1 1 NRR/DBRO ENRIT 1 1 NRR/DSRO/EIB 1 1 NRR/DBRQ/RBIB 1 01 <<. 1 1 EXTERNAL: LPDR 1 1 NRC PDR
'NSIC 1 TOTAL NUNBER OF CQPIEB REQUIRED: LT1R 2V ENCL 28
0 li k
I 4
p U
1MIAO RA O MaHawX NIAGARAMOHAWKPOWER CORPORATION/301 PLAINFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474-1511 March 18, 1987 (NMP2L 1007)
U ~ S. Nuclear Regulatory Commission Attn: Document Control Desk Hashington, D. C. 20555 RE: Nine Mile Point Unit 2 Docket No. 50-410 Gentlemen:
As discussed with the Nuclear Regulatory Commission staff on March 18, 1987, Niagara Mohawk Power Corporation requests that the Nuclear Regulatory Commission accept an alternate to Regulatory Guide 1.96, "Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Hater Nuclear Power Plants," for Nine Mile Point Unit 2. The alternate position is discussed in this letter and Attachment A. In summary, Niagara Mohawk has performed an analysis that demonstrates that a leakage control system is not required.
This conclusion is based on the results of a radiological analysis, comparison of the Unit 2 design features to those identified in NUREG 1169, and the conservatism of the main steam isolation valve leakage limits determined by the design basis accident radiological assessment.
On January 14, 1987 Niagara Mohawk submitted a radiological analysis (enclosed as Attachment B) for Unit 2. The analysis demonstrates that with up to 150 scfh leakage through the main 'steam lines, using the methodology contained in NUREG 1169, the offsite doses and control room habitability remain within regulatory requirements. This* conclusion is independent of the type of isolation valve.
Niagara Mohawk has performed an assessment comparing the Unit 2 design to the design in NUREG 1169, "Technical Findings 'Related to Generic Issue C-8, Boiling Hater Reactor Main Steam Isolation Valve Leakage Treatment Methods."
The assessment makes the following significant conclusions.
- l. Unit 2 critical plant parameters governing leakage of radioactivity are comparable to that described in NUREG 1169. Therefore, the method and results described in this NUREG are applicable.
- 2. Unit 2 has the means in place to effectively address main steam isolation valve leakage without a leakage control system.
8703200296 870318
'PDR P "
ADOCK 050004i0 PDR
~I 11 1
Leakage Control Sp em Page 2
- 3. As noted above, the offsite and personnel doses have been shown to be acceptable with main steam isolation valve leakage up to 150 scfh without a leakage control system. Thus, there is no necessity to install a main steam isolation valve leakage control system.
The Discussion Section of Regulatory Guide 1.96 indicates that the Nuclear Regulatory Commission was concerned with main steam isolation valve leakage during a Design Bases LOCA. Nuclear Regulatory Commission staff studies at other plants have shown that ma'in steam isolation valve leakage up to the technical specification value would exceed the 10CFR100 guidelines.
However, the Unit 2 design basis accident radiological assessment submitted in Section 15.6.5 and Table 6.2-55(a) of the Final Safety Analysis Report included 6 scfh of main steam isolation valve leakage (for each line) as part of secondary containment bypass leakage. The 6 scfh leakage value is contained in the Technical Specifications, and must be confirmed periodically by testing as required by the Surveillance Requirements of the Technical Specifications.
The Nuclear Regulatory Commission, in Supplement 5 to the Safety Evalua-tion Report, approved the leakage values in the Technical Specifications.
Page 6-44 of the Safety Evaluation Report also indicated that a leakage control system was not required for Unit 2 to meet Regulatory Guide 1.96.
While the main steam isolation valves are being changed to wye pattern globe valves, this modification does not change the results of either of the radiological analyses described above.
Therefore,- Niagara Mohawk Power Corporation intends to implement the alternate approach to Regulatory Guide 1.96 described in Attachment A. We request that the staff expeditiously review this matter and notify Niagara Mohawk of the acceptability of this approach as soon as practicable. We stand ready to meet with you to discuss this matter at any time.
Very truly yours, NIAGARA MOHAWK POWER CORPORATION C. V. Mang Senior Vice President CVM/bwr (2S44G) cc: Regional Administrator, Region I Ms, E. G. Adensam, Project Director Mr. W. A. Cook, Resident Inspector NMP2 SSC File (2)
Project File (2)
ATTACHMENT A Regulatory Guide 1.96 Alternate Position to i8/8y 8703200296.
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F Nine Milo Point Unit 2 FSAR TABLE 1.8-1 (Cont)
Re ulator Guide 1.96 Revision 1 June 1976 Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Planta FSAR Sections 1.2.9.11, 5.4.5, 6.2.3.2.3, 15.6.5 Position Main steam isolation valve (MSIV) leakage, at the maximum rate allowed by the Technical Specifications, has been included in the secondary containment bypass leakage analysis (Section 6.2.3,2.3)'and in the LOCA radiological consequence analysis (Section 15.6.5). These design-basis analyses demonstrate that the calculated exposures are within the guidelines of 10CFR100 and 10CFR50, Appendix A, General Design Criteria 19.
In addition, a qualitative comparison has been made between Unit 2 and the plant used as the basis for analyses presented in NUREG-1169, "Technical Findings Related to Generic Issue C-8; Boiling Hater Reactor Main Steam Isolation Valve Leakage and Leakage Treatment Methods. "
This comparison demonstrated that the design features of Unit 2 are sufficiently similar to the NUREG-1169 base plant, such that the conclusions of NUREG-1169 are considered directly applicable to Unit 2, NUREG-1169 concluded that the overall risks from the accident sequences in which MSIV leakage could be a significant factor are low without a leakage control system, and alternate fission product handling techniques, which make use of the holdup volume of the main steamlines and condenser, produce significant reductions in offsite dose consequences.
therefore concluded that an MSIV leakage control system is It is not required for Unit 2.
Amendment 110 of 169
r E I. EVALUATION OF NINE MILE POINT UNIT 2 UTILIZING NUREG 1169 AD INTRODUCTION AND
SUMMARY
OF RESULTS The Nuclear Regulatory Commission established a review team to address main steam isolation valve (MSIV) leakage. Their review concluded that "the overall risks from the accident (LOCA) sequences in which MSIV leakage is a significant factor are low ... and alternate management schemes produce significant dose reductions."
To assess the ability of the globe valves to meet the safety requirements related to isolation and leakage control, Niagara Mohawk has performed a comparison of the Unit 2 design to the design evaluated in the Nuclear Regulatory Commission NUREG 1169. The comparison shows that the results of this study are applicable to Nine Mile Point Unit 2. This report provides justification that a MSIV leakage control system is not required. Section II of this report provides a comparison of Unit 2 to the NUREG 1169 reference plant; Section III evaluates the NUREG 1169 Leakage Control methods; Section IV provides a summary of the radiological dose assessment; and Section V contains the conclusions.
As stated in the document, the purpose of NUREG 1169 was to determine: 1) the adequacy of . industry efforts to identify
. and correct causes of excessive MSIV leakage, 2) the basis for any change in the allowable MSIV leakage rate, 3) the need for a safety-grade Leakage Control System (LCS), and 4) the specific areas of regulations and guidance that may be necessary to implement the findings.
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The approach used in NUREG 1169 was to evaluate the effects of MSIV leakage in terms of offsite doses following a LOCA, using realistic assumptions concerning the equipment, facilities and site character-i sties available to mitigate the effects of a LOCA.
NUREG 1169 indicated that alternate treatment methods (discussed in the following sections as they apply to Unit 2) are highly effective in reducing the radioactive material released such that the MSIV leak rate could be increased significantly without exceeding any dose limitations. Further, the NUREG indicates that the reliability of the leakage control system is actually lower than some of the alternate treatment methods. Therefore, the installation of LCS is unnecessary and higher MSIV leakage rates can be adequately mitigated by existing alternate treatment methods.
This report demonstrates that the physical layout and design of Unit 2 is such that the health and safety of the public are assured even were the Unit 2 MSIVs to leak in excess of their Technical Specifica-tion surveillance limits. This was demonstrated by comparing the reference plant in NUREG 1169 (WNP-2) with NMP2 and by a radiological calculation that shows the results are acceptable. This calculation is provided as Attachment B. The calculation shows that a total leakage for all main steam lines of up to 150 scfh (or 500 scfh with beta shielding) would not result in control room doses exceeding regulatory requirements specified in 10CFR50 Appendix A, General Design Criteria 19; over 1000 scfh/line leakage would be necessary to exceed 10CFR100 offsite guideline values to the public.
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I I. COMPARISON OF PLANT PARAMETERS TO NUREG 1169 This section provides the comparison between the reference plant (WNP2) and NMP2. The significant points of comparison are discussed below:
Table I shows the comparison of the -critical parameters affecting the radiological consequences of a LOCA. Tables II through VI show a detailed comparison between the reference plant for NUREG 1169 analysis and NMP2. The parameters governing the progression of a LOCA are the thermal power, thermal-hydraulic design, and the ECCs. Review of this data in Tables I, II, and III shows nearly identical design for both plants. Therefore, the expected source-term for NMP2 would be identical to NUREG 1169.
NMP2 steam lines. are expected to cooldown at a somewhat higher rate than the reference plant, primarily due to higher thermal conductivity of the insulation and larger number of supports.
This is expected to enhance particulate removal due to thermo-phoretic aerosol deposition.
Any leakage through the paths described in NUREG 1169 considering the Unit .2 design would result in the radioactive releases being release'd either from the main stack or from the turbine building (dispersion factors are represented by the radwaste/reactor build-ing (RW/RB) vent in Table I). In either case, the dispersion factors are comparable to those used in the NUREG 1169 analysis.
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Since Unit 2 and the reference plant have similar designs and systems, the NUREG 1169 probabi 1 i sti c risk assessment scenario i s judged to be equally applicable, In addition, NMP2 has an auxiliary boiler steam supply system to sIjpport operation of the Steam Jet Air Ejector Offgas pathway for MSIV leakage control (after isolation of the condenser from the reactor). This path is highly effective in removing and delaying radioactivity prior to release, and would reduce the releases identified in Section IV. Because the releases would be reduced, the risk would also be reduced.
EVALUATION OF LEAKAGE CONTROL METHOD This section compares the alternate leakage treatment methods discussed in NUREG 1169 to the design of the Nine Mile Point Nuclear Station Unit
- 2. The alternate leakage treatment methods contained in NUREG 1169 are:
Isolated Condenser Mechanical Vacuum Pumps Steam Jet Air Ejectors-Offgas System Isolated Steam Lines Each of these are discussed in relation to NMP2 as follows:
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A. ISOLATED CONDENSER This leakage treatment method takes advantage of the main condenser to hold up the release of fission products from the main steam isola-tion valves (MSIV) and the main steam lines (MSL). The condenser is isolated from the turbine building and the outside environment. The isolated condenser method addressed in NUREG 1169 has separate paths to convey leakage to the condenser, The first path (using the turbine bypass valves) requires operator action to open the bypass valves. The second path uses the main steam drain valves in lieu of the turbine bypass valves to connect to the condenser. NUREG 1169 mentions that in some plants, the steam line drains are of a Fail-open design, requiring no operator action.
The Nine Mile Point Unit 2 main steam line drains automatically open on loss of air power, and first stage turbine pressure, Both Unit 2 and the reference plant discussed in NUREG 1169, use this completely passive system; the main steam lines communicate with the condenser without operator action. It is also possible to connect the main steam lines to the condenser by way of the turbine bypass valves, but this would require operator action to initiate the turbine electrohydraulic (EHC) system. In either case, the NMP2 isolated condenser leakage will migrate through the low pressure turbine seals into the turbine building and into the environment as described in NUREG 1169.
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0 B. MECHANICAL VACUUM PUMPS NUREG 1169 addresses the use of both condenser mechanical vacuum pumps and the gland seal and exhaust system blowers after an acci-dent. By use of this equipment, the condenser is- kept at a lower pressure than the surrounding environment. Thus, MSIV leakage will migrate through the main steam line to the condenser, assuming that the MSL drains are open and/or the turbine bypass valves are open.
The mechanical vacuum pumps or gland seal exhausters will discharge the leakage to the stack (elevated release point). The NMP2 system design is the same as the described in the NUREG.
C. STEAM JET AIR EJECTOR OFFGAS SYSTEM NUREG 1169 describes a highly desirable mode of operation for control of MSIV leakage which uses the plant's existing steam jet air ejec-tors, steam seal system, and off-gas system to first collect the leakage and then discharge this leakage through the off-gas system where it is filtered, treated, and delayed. In addition, the discharge from the off-gas system is then sent to the stack (elevated release point).
The NMP2 installation meets all the recommendations for the most desirable steam jet air ejector-offgas system operation. The NMP2 design incorporates two electric boilers, either one of which can produce a sufficient amount of steam to re-establish operation of the 2844G
steam jet air ejectors and the gland and exhaust system if offsite power is available. Therefore, the Unit 2 design can accom-plish necessary filtering and delay of the radioactive gases.
D. ISOLATED STEAM LINES NUREG 1169 evaluates the condition in which the main steam lines are sealed off from the condenser, turbine, and environment. The main steam lines become a cavity to isolate the MSIV leakage from the environment. In this case, the turbine stop and control valves and bypass valves can pass some leakage which will eventually migrate to the environment. The NMP2 installation is again as described. in the NUREG. However, this scenario has a low probability since the main steam drains automatically open upon loss of air or power.
E. PROBAB I L I ST I C R I SK ASSESSMENT NUREG 1169 contains a probabilistic analysis of each path including leakage control system pathway. Comparing the NMP2 plant with the analysis contained in the NUREG indicates that there are few differ-ences in the NMP2 installation except as noted below. These differ-ences resul.t in lower releases and risks at Unit 2 than the reference plant.
If one of the Auxiliary Boilers were in operation prior to the LOCA, steam would be immediately available (as long as power i s available). If one of the Auxiliary Boilers was not in operation prior to the LOCA, a delay (as long a 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) in steam availability could occur.
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- l. Isolated Steam Line Flow Path This condition has a low probability at the NMP2 plant due to the fact that the main steam line drains automatically open on loss of air, power, or turbine first stage pressure. The more likely path is the isolated condenser.
- 2. OFF-GAS SYSTEM AVAILABILITY The probability tree for the system using the steam jet air ejectors with the off-gas system and the steam seal and exhaust system is actually better for NMP2 because of higher avai la-bility than that contained in the NUREG. This improvement is described below:
- a. Two electric boilers are installed, each of which can supply 40,500 lb. of steam per hour. Nith either boiler, the steam jet air ejectors, the steam seal and exhaust system, and the off-gas system can be main-tained through the event.
- b. The main steam line drains are a passive system; thus increasing their. probability of opening during a LOCA.
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c ~ In the event of a Loss of Offsite Power (LOOP) and LOCA with the NMP2 installation, the condenser vacuum, steam jet air ejectors, off-gas system, and gland seal exhaust system can be re-established once power is restored.
IV. RADIOLOGICAL ANALYSIS IN CONFORMANCE WITH NUREG 1169 Based on the NUREG 1169 comparison with NMP2, the isolated condenser path is the conservative scenario for calculation of radiological conse-quences. The allowable HSIV leakage was determined based upon a simpli-fied main condenser model for the beta skin and whole body gamma doses, while a direct comparison ratio method was used to determine the thyroid doses. The beta and gamma dose evaluation model utilized holdup of the HSIV leakage in the main condenser and subsequent release of radioactiv-itiy to the environment. No credit was taken for holdup of noble gases in the main steam lines, drain lines or turbine building. Additionally, the volume reduction due to steam condensing in the piping or components prior to being released was not considered. The above conservative analysis provided results indicating that the most restrictive radiation limit was the Control Room beta dose. The analysis demonstrates that a HSIV leak rate of 150 scfh total for all steam lines would not result in personnel doses in excess of regulatory limits. The maximum leakage rate could be increased to 500 scfh from the main steam lines with appropriate beta shielding (such as overalls) and meet control room habitability guidelines and a value as high as 1000 scfh/line would not produce 2844G
offsite doses in excess of regulatory requirements, The calculation of radiological impact is provided in Attachment B.
V. FINAL
SUMMARY
OF COMPARISON TO NUREG 1169 NMP2 has in place what NUREG 1169 characterizes as, "The Highly Desirable Mode of Operation," that is, a method of collecting, treating, and discharging from the stack all leakage from the main steam isolation valves. This is accomplished by:
- a. A passive steam line drain system.
- b. Electric boilers capable of providing steam to the steam jet air ejectors, off-gas system, and turbine gland seal and exhaust system.
c ~ In the event of a LOCA and/or LOOP, NMP2 has the capability to re-establish the condenser vacuum, the operation of the steam jet air ejector, the operation of the gland seal and exhaust system and the offgas steam once off site power is restored.
Following simultaneous LOCA and LOOP, the NMP2 plant would automatic-ally align itself in the condition defined as Isolated Condenser Pathway. It is unlikely that an isolated steam line pathway, a somewhat less effective method to control MSIV leakage, would occur at NMP2.
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The following conclusions are drawn from this analysis:
- 1. NMP2 has the capability to effectively control the MSIV leakage in a manner similar to NUREG 1169, without a leakage control system.
- 2. NMP2 meets the NUREG 1169 description of the most desirable operating mode "steam jet air ejector offgas" which is available following a LOCA or temporary loss of off site power.
- 3. NMP2 main steam lines are expected to cool down at a higher rate, thus, further reducing off site dose.
- 4. The radiological analysis has determined that leakage up to 150 scfh for all four mainsteam lines (or 500 scfh with beta shield-ing) would not result in personnel doses in excess of regulatory limits, or offsite doses in excess of regulatory guideline valves.
Therefore, Niagara Mohawk Power Corporation concludes that a MSIV leakage control system is not required.
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TABLE I RADIOLOGICAL CONSEQUENCE PA~~<TERS PARAMETER NMP2 NUREG-f169 Plant'Type BWR-5 BWR-5 Mark II Mark II Pover (MWth) 3323 3323 Combined Technical Specification <0.19 0.27 Leakage Rate for MSIVs,volume percent/day MSIV Leakage Pathways* o Isolated o Isolated Condenser Condenser o Isolated Steam o Isolated Steam Lines Lines o Mechanical o Mechanical Vacuum Pumps Vacuum Pumps o SJAE - Offgas o SJAE - Offgas System System Steam Line Details Pipe Size 28 II 30ll Wall Thickness 1 2/8" 1 3/8" Insulation Thickness 4tt pit Insulation Thermal 0.03 0.02 Conductivity, Btu/ft-hr'F Number of Pipe Supports 36+ 29 Support Spacing 1'in 8'in 25'ax 35'ax
- Section III provides detailed discussion
+ These supports are typical of one line~ supports for headers and valves are not included Page 1 of 2
TABLE I RADIOLOGTCAL CONSEQUENCE PARAlKTERS PARAMETER NHP2 NURZG-j?69 Turbine System Supplier GE Condenser Volume, ft3 123s000 120,000 Condenser Horizontal 214,000 252, 000 Deposition Area, ft2 Dispersion Factors+ RWIRB Vent Stack EAB 0-2hr 19.00xlO"5 2.97x10 7.50x10 5 LPZ 0-8hr 1. 78xlO" 5 1.03x10"5 2.80x10 5 8-24hr 11.90xlO 6 0.88xlO 6 3.45xlO 6 1-4 days 4.93x10-6 0. 37x10-6 1.59xl0 6 4-30 days 1.40xlo 6 0.10x10 6 1,02x10 6
- NUREG-1169 data taken from the WNP2 FSAR Chapter 15. This data have been used for a variety of release points in WNP2 FSAR.
Page 2 of 2
TABLE NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS NMP2 NURZC-1169 THERMAL AiiD HYDRAULIC DESIGN Design powe'r, MWt (ECCS design basis) 3,463 3,468 Steam flov rate, millions lb/hr 14.27 14.30 Core coolant flov rate, millions lb/hr 108.5 108.5 Feedwater flov rate, millions lb/hr 14. 56 14.26 System pressure, nominal in steam dome, psia 1,020 1,020 Average po~er density, kW/1 49.15 49. 15 Minimum cri,tical pover flux ratio (MCPR) 1. 24 1.24 Coolant enthalpy at core inlet, Btu/lb 527.5 527 '
Core max exit voids within assemblies 76.2 76 Core average exit quality, X steam 13.10 13.5 Feedva'ter temperature, 'F 420 420 Desi n Power Peakin Factor Maximum relative assembly power 1.40 1.40 Axial peaking factor 1.40 1.40 REACTOR UESSEL DESIGN Mater ia 1 Lov-alloy steel/ Carbon steel/
stainless clad stainless clad Minimum base metal thickness (cylindrical section), in 6.1875 6.75 Minimum cladding thickness, in 1/8 1/8 Design pressure, psig 1,250 1,250 Design termperature, 'F 575 575 Inside diameter, ft"in 20"11 20-11 Inside height, ft-in 72-5 72"11 Page 1 of 2
TABLE
&P2 NUREG-LI69 REACTOR COOLANT RECIRCULATiON DESIGN No. recirculation loops Design pressure Inlet leg, psig 1,250 1,250 Outlet leg, psig l,eso(1) l,eso( >>
1,550(2) 1,SSO(>>
Design temperature, F 575 575 Pipe diameter, in 24 24 Pipe material, AISI 316K 304 j316 Recirculation pump flow rate, gpm 47,200 47,250 No. jet pumps in reactor 20 20 HAIN STEAM LINES No. steam lines 4 4 Design pressure, psig 1)250 1,250 Design temp. 'F 575 575 Pipe material Carbon steel Carbon steel (1) Pump and discharge piping to and including the discharge block valve.
(2) Discharge piping from discharge block valve to vessel.
Page 2 of 2
TABLE III EHERGENCY CORE COOLING SYSTEM DESIGN CHARACTERISTICS NMP2 NUREG-1159 Low Pressure Core S ra S stem No ~ loops 1 1 Flow rate, cpm 6,350 9 128 psid 6,250 8 122 paid Hi h Pressure Core S ra S stem No. loops 1 1 Flow rate, cpm 1,550 9 1,130 psid lc650 8 1<13.0 psjd 6,350 9 200 psid 6,250 9 200 psid Automatic De ressurixation S stem No. systems No. relief valves Low Pressure Coolant In ection No. loops 3 3 Flow rate, gpm/pump 7,<50 8 26 paid 7;450 8 20 paid
TABLE IV CONTAIQKNT DESIGN CHARACTERISTICS NMP2 NUREG-li69 Primar conte'inment(i)
Type Over & under Over & under pressure pressure suppression suppression Mark II Mark II Construction Reinforced Steel free concrete standing steel liner Drywe 1 1 Frustum of cone, Frustum of cone, upper portion upper portion Pressure suppression Cylindrical Cylindrical chamber lower portion lower portion with elliptical bottom Pressure suppression chamber- 45 internal design pressure, psig Prcssure suppression chamber " 4.7 external design pressure, psig Drywell - internal design 45 45 pressure, psig Drywell - external design 4.7 pressure, psig
\
Drywell fre>> volume, ft3 303,418 200,540(2)
Pressure suppression chamber 192,028 144,184 free volume (min), ft3 Pressure suppression pool 154,794(4) 112,177 water volume (max), ft3 Page 1 of 3
TABLc IV PitP 2 NUR"G 1'.69 Submergence of vent pipe below 9,5 min LL.67 min suppression pool surface, ft LL.O max 12.00 max Design environmental temperature 340 340 of drywell, F Design envi.onmental temperature 270 275 of pressure suppression chamber, 'F Downcomer vent pipe L.37<5~ 1.9 pressure loss factor Break area/total vent area 0.0108 0.01,05 Calculated maximum pressure 39.7 34,7 after blowdown to drywell, psig Calculated maximum pressure 34.0 28.0 in suppression chamber, psig CalcuLated maximum initial 50 35
,pressure suppression pool temperature rise, 'F Leakage rate, 'X free volume/day 0.5 at 45 psig Reactor Buildin Type Controlled Controlled leakage, Leakage, elevated elevated release<6> release Cons true t ion Lower levels Reinforced Reinforced concrete concrete Upper levels Steel super- Steel suoer-structure and structure and s iding siding Roof Steel decking SteeL decking Page 2 of 3
TABLE IV NMP2 NUREC-1'.69 Internal design pressure, psig 0.25 0,25 Design inleakage rate, 100 100
- 7. free.'volume/day at 0.25 in H20 (1) Where applicable, containment parameters are based on design power.'2)
Maximum water in suppression pool.
(3) Does not include water in the pedestal.
(4) At high water level.
(5) Includes entrance and pipe friction.
(6) For accident conditions.
Page 3 of 3
1 TABLE V ELECTRECAL POWER SYSTEM DESIGN CHARACT:.R STECS
&P2 NUREC-1169 Offsice Power S stem Outgoing lines (No."rating) 1-345kv 1-500kv Encoming lines (No. -ra t ing) 2-115kv 1-230kv 1-115kv Onsite ac Power S stem Normal station service t r ans forme r s Reserve station service 3(1) transformers Standby diesel generators 3(2) 3(2) 4,160V ESF buses 3(2) 3(2)
ESF buses 3-600-V<2) 3-4go-V<2) dc Power Su 1 4-24V Batteries (No.-volts) 6-1.25V<3) 5-125v<3) 4-24V 1-250V 2-24V Buses (No.-volts) 6-125-V<3) 5-125v<>>
2-24 V 1-250V (1) Includes one auxiliary boiler transformer.
(2) Includes an HPCS diesel generator.
(3) KPCS battery and bus included.
0 TABLE VT POWER CONVERSZON SYSTEM DESIGN CHARACTERISTICS NMP2 NUREC-ii))
Design power; NWt 3,463 3,468 Design pover, MWe, gross 1)202 1, 205 Generator speed, RPM 1.) 800 1,800 Design steam flow, lb/hr 14 3 106 l5.0 x 106 Turbine inlet pressure, psia 965 970 Turbine B ass S stem Capacity, percent of turbine 25 25 design steam flov Main Condenser Heat removal capacity, Btu/hr 7,830 x 1P6 7,702 x 106 Ci.rculatin Water S stem No. Pumps 6 8 Flov rate, gpm/pump 105,000 82)000 Condensate and Feedvater S stems Design flov rate, lb/hr 14.917 x 1P6 14.260 x 106
.No. condensate pumps 2 running, 1 stdby. 3 running No. condensate booster pumps 2 running, 1 stdby. 3 running No. feedvater pumps 2 running, 1 stdby. 2 running Condensate pump drive ac pover ac power Condensate booster pump drive ac pover'c ac pover Feedvater pump drive paver Turbi.ne Heater drain pumps 3 running
ATTACHMENT B RADIOLOGICAL ANALYSIS OF MSIV LEAKAGE 2844G
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