LD-96-021, Submits Changes to Sys 80+ Std Plant Design Control Document as Requested by for Staff Review & Approval
| ML20117K780 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 06/11/1996 |
| From: | Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LD-96-021, LD-96-21, NUDOCS 9606120221 | |
| Download: ML20117K780 (68) | |
Text
{{#Wiki_filter:. . Jk D D Mpp l l June 11,1996 ! LD-96-021 l Docket 52-002 Attention: Document Control Desk l U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Subject:
Changes to the System 80+* Standard Plant Design
Reference:
Letter, B. K. Grimes to C. B. Brinkman," Preparation of Final Design Control l Document for the System 80+ Standard Plant Design," 5/15/96 ) l Dear Sirs l The purpose of this letter is to submit the attached draft changes to the System 80+ Design Control l Document (DCD), as requested in the Reference letter, for Staff review and approval. These design ! changes are itemized on the attached table with details shown on the attached marked-up DCD pages; these changes were discussed with the Staff on June 6,1996. Approved changes will be documented by printed replacement DCD pages containing a margin bar l adjacent to each change and a revision date of 6/96. All such margin bars and revision dates will be i deleted from the DCD when it is reprinted following design certification rulemaking. Please call me, Stan Ritterbusch (860-285-5206) or Virgil Paggen (860-285-4700) if you have any questions. Very truly yours, COMBUSTION ENGINEERING, INC.
&l.. n C. B. Brinkman Director, Nuclear Licensing cc: J. N.. Wilson (NRC)
S. M. Franks w/o enclosure (DOE) } Attachments: As stated 9606120221 960611 npI PDR ADOCK 05200002 04 :J r i e PDR ABB Combustion Engineering Nuclear Systems Carbustion Erginoenng. Inc. 2000 Day HJI Road Telephone 1860) 6881911 PO Bas 500 Fax (600) 265203 Windsor CT 000E0500
Patenti:1 System 80+ Design Ch ngea r item Ref DCD I No. Section Design Change Description Tbl 1.8-7,Tbl in ASME Section XI Code has been revised to ehminate inservice hydrostatic testing. Code now requires shop test i 3 9; and initial in-plant test at 125% of design pressure, while remaining in-plant leak-rate tests would be conducted at Ch 5,16 100% of design pressure. (Change is approved by NRC). Code Cases N-453, N-498 address issue. No impact on Ch6 ITAAC or FSER. Action; modify P T curves to remove " Inservice test"line, add Code Case N-498. ( 2 Sect 6.5.3.4, insufficient net positive auction head is available to the S80+ shutdown cooling pumps when aligned for t Figures 6.3.2 containment spray due to frictionalloss in crossover piping. Increased SCS piping & valve sizes; reduce allowable 1A & 6 3.2-1B. containment spray runout flow to provide reqd NPSH when used for Ctmt spray function. l 3 Tbl 3.2-1 Correct the seismic classification, quahty class of the vacuum breakers, the piping from the PSVs to the IRWST, and spargers; correct DG heading (pg 3.2-13), correct potable water heading (pg 3.2-18). l 4 Ch 16. Fig Evaluate need to revise Tech Spec on IRWST temperature. Fig 3.5.41 indicates IRWST hmft is 40F; analysis hmit 3.45-1 is 60F. Confirm whether Tech Spec Figure in Ch 16 should be updated to 60F. 5 Ch 4; Twelve (12) CEA locations in core center region should be added and designated as " spares." These would allow Figs in Ch 3,5 addrtional maneuvenng capabihty. Requires correction of CEA locations shown in ITAAC Fig 2.2.13, ADM Fig 4.2-
- ITAAC 11, ADM Fig 4.3-46, ADM Fig 4.3-47. Also modify Section 4.2.2.4 to accommodato extra CEAs.
6 Ch5 Revise mid-loop level monitor to accommodate ex-vessel device. 7 Ch 6, Ch 3.6 Reduce SDS line/ valve size from 6" to 4" per YGN quahfication data. Revisa Figure 5.1.2-3. 8 Ch3 Modify Piping Evaluation Diagrams to be consistent with NRC-approved criteria for AP600. Change load combination factor from 1.4 to 1.0. 9 Pg 9.3-29 Section 9.3.4.2.1 states interlock provided so only one chg pump can be operated during all modes of plant operation. [Ch 15 limits chg flow to 150 gpm for boron dilution; also Tier 1 commitment.] This means that one charging pump must be stopped before the standby pump can be started. Momentary loss of RCP seat injection may occur. 10 Ch3 Increase damping for envelope response spectrum analysis of piping to 5% to be consistent with NRC-approved criteria for AP-600. l 11 Tbl 3.6-3 Delete PSV rehef lines, items 40.. 43 since valves mounted directly on pzr; correct items 58, 59. 12 Tbl 3.9-2 Correct Table title to include component supports in loading combinations.- 13 Tbl 4.2-3 Add ANO inspection program scope for year 1989/ cycle 7. 14 Tbl 5.4.7-2 Correct failure mode entry to " fait closed" for shutdown cooling pump dischg isolation valve. , l 15 Pg 9.3-30 Correct statement regarding fluid retum to RCS when in shutdown coohng mode. 16 Pg 9.3-37 Revise CVCS system " redundancy" statements to be consistent. 17 Tbl 9.3.4-4 Change VCT normal operating pressure to be 20-50 psig. i 18 Pg 10.3-7,13 Revise MSIV bypass valve closing time to be 5 seconds or less, rather than 10 seconds. 5 seconds was used in 14 safety analysis. 19 Ch 19.7.5 Delete Tables 19.7.5.1-1 through 19.7.5.4-7. These tables should be htentionally blank. 20 Pg 19.11-145 Revise statement re: core uncovery at 7700 seconds to be consistenEh' red. ed Fig & Table. l i ! File cMedWed-dchg. doc Date: June 11,1996 Page1 l i ,
System 80+ DCD - Potential Design Changes Item Number: 1 Summary Descrintion: Delete Inservice RCS Hydrostatic Testing Affected DCD Section(s): CDM: None , ADM: Table 3.9-1 Chapters 5 and 6 , l Description of Change: 1 A recent ASME B&PV Cod Case has been published which provides an alternative to performing the 10-year hydrostatic pressure test. Code Case N-498,"Altemate Rules for 10 Year Hydrostatic Pressure Testing for Class 1 and 2 Systems Section XI, Division 1", indicates that a system leakage test can be conducted at or near the end of each inspection period , prior to reactor startup instead of the Hydrostatic Pressure Test. This code case was approved by the NRC as indicated in Reg Guide 1.147, Rev 09. Currently the DCD (Table 3.9-1) specifies that , 15 RCS and secondary hydrostatic tests are included in the stress analysis of Code Class 1 and CS components. This number will be reduced from 15 to 10 occurrences during the plant life time. The RCS and secondary leak tests have adequate margin in the number of occurrences listed in Table 3.9-1 to account for the additional leak tests which will be performed instead of the hydrostatic pressure tests. Table 1.8-7. "ASME Section III Code Cases applicable to System 80+ will be revised to included Code Case N-498. In addition, Section 5.2.4.6 of the DCD will be revised to include reference to the Code Case. DCD Markups Attached? Yes 1 TKSdedch. doc 1 6/11/96
l Svstem 80+ Ooston controt Documen: Table 3.9-1 Transients Used in Stress Analysis of Code Class 1 and CS Components O (Cont'd.) Test Conditionst21 Occurrencest21
- 1. RCS hydrostatic test
[-[Prunary pressure cycles from atmospheric to 3125 / D, psia ah
------_;._:.% between 120 and 400'FR _f
- 2. RCS leak test 200.
-rf ^ .
[Prunary pressure cycles from atmospheric to 2250 psia at a\.
- p.uww between_1.2_0 and 400'F1
- 3. Secondary hydrostatic test m g_
[ Secondary pressure cycles from atmospheric to 1500 psia at a /0, bi""" !""""'ature of.190*F1 . __ _ _ ( 4. Secondary leak test ( [ 200. 7 - [ Secondary pressure cyclea from atmospheric to 1200 psia at a 3
==!- -
--- '-5 _ww of 200'F1 -
- 5. SIS /SCS check valve operability test 500.
- 6. SIS /SCS preoperational and maintenance test 240.
O Upset Conditions!!! OccurrencesI71
- 1. Decrease in feedwater twswe 20.
- 2. Increase in feedwater flow rate 20.
- 3. Increase in steem flow rate 20.
- 4. Inadvertent openmg of a steam generator relief or safety valve 10.
- 5. Ims: of load (turbine speed control system operates normally) 19.
[ loss of electrical load and normal turbine / generator runbeck to house load]
- 6. Turbine trip 20.
l .
- 7. Iess of candansar vacuum 20.
l 8. Imss of non4mergency AC power to the station =mhanes 10. l
- 9. Iees of normal feedwater flow 20.
- s
, [ Subsequent actuation and cycling of cold emergency feedwater to the steem generators] { } 10. less of forced reactor coolant flow 20. j 11. Uncontrolled CEA withdrawal from subernical or low power condition 10. O I Sp-eno asmeest oesy,ersac esp. J.sJo l
]
. Svstem 80+ o sian controloocarumnt Table 1.8-7 ASME Section III Code Cases Applicable to System 80+
- O Case Title N-4-11 [1337-11] Special Type 403 Modified Forgings or Bars, Class 1 and CS; 7/13/87.
N-60-4 Material for Core Support Structures 7/27/88. N-71-15 Additional Matenals for Subsection NF, Classes 1,2,3 and MC Cnmanaaa' Supports Fabricated by Welding: 12/16/89. N-1221 Evaluation of the Design of Rectangular Cross-Section Attachments on Class-1 Piping; , 7/27/92. N-192-2 Use of Braided Fleaible Connectors, Class 2 and 3; 9/17/87. N-247 Certified Design Report Summary for Component Standard Support. Class 1,2,3, and MC;
- 1/21/88.
N-249-10 Additional Materials for Subsection NF, Classes 1,2,3 and MC Component Supports Fabricated without Welding: 5/06/89. N-262 Resistance Spot Welding for Structural Use in Component Suppons: 7/28/88. N-284 Metal Containment Shell Buckling Design Methods; Section III, Division 1, Class MC; 8/25/80. N-309-1 Identification of Matenal for Component Supports; 7/28/88. N-313 Alternate Rules for half-Coupling Branch Connections, Class 2; 11/28/86. N 318-4 Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 or 3 Piping; 12/11/89. N-319-1 Evaluation of Stresses in Butt Welded Elbows for Class 1 Piping: 7/24/89. , N-391-1 Evaluation of the Design of Hollow Circular Cross Section Welded Attachments on Class 1 Papag; 7/24/89. N 392-1 Evaluatson of the Design of Hollow Circular Cross Section Welded Attachmanta on Class 2 and 3 Pipings: 12/11/89. N 393 Repaar Welding Structural Steel Rolled Shapes and Plates for Components Supports; 7/30/89. N-411-1 Alternative Damping Values for Response Spectra Analysis for Class 1,2, and 3 Pipmg; 2/20/89. N-420 linear Energy Absorbmg Supports for Subsection NF, Class 1,2, and 3 Construction; 2/14/88 N-430 Alternative Requu=maata for Welding Workmanship and Visual Acceptance Criteria for
. Class 1,2,3, and MC Linear-type and Standard Suppons: 2/28/89.
N-433 Non-threaded Fasteners for Class 1,2, and 3 e'aa aaa s Piping Supports: 12/16/89. N-4741 Design Stress Intensities and Yield Strength Values for UNS NO6690 with a mmunum specific Yield Strength of 35 kai, Class 1 Cv- 3/05/90. N-476 Class 1,2,3, and MC Linear C+4= - Supports - Design Criteria for Single Angle Mesnbers, Subsection NF; 5/06/89. H Ljqq AHenaVtJe Lus for to-year Hydros taHc Pa ssxx Test m er class i asd I, 5ss tems Sevnos c, O D b 5 'iO r4 I 4pesed aes(pn assesser- emeshesim nye f.s.40
Svstem 80+ Desian ControlDocument n 't If 5.2.4.4 Inspection Intervals v
.g k The IWA-2000 examination program for the 120 month inspection interval will be defined in the ISI plan.
gJ ~ The ISI Plan for all Code Class I systems and components will be in accordance with the ASME Code g g Section XI edition in effect per 10 CFR 50.55a 12 months prior to the issuance of an operating license (initial interval). W'5
-d O
f 5.2.4.5 Evaluation of Examination Results V t Evaluation of c?nmination results for Class 1 components will be conducted in accordance with Articles
\y d IWA-3000 and IWB-3000 of ASME Section XI.
e3 Unacceptable indications will be repaired in accordance with the requirements of Articles IWA-4000 and (3y IWB-4000 oi ASME Section XI. Criteria for establishing need for repair or replacement shall be per IWB-3000, w- ,
#Tvg{ [ 5.2.4.6 System IAakage and Hydrostatic Tests j
)
,4 j ^ Ibe hydrostatic and system leak tests for the reactor pressure vessel and reactor coolant pressure
)
a f boundary will be conducted in accordance with the requirements of Articles IWA-5000 and IWB-5000 of ASME Section F==in= tens performed during these tests will be corAucted without the removal 0g,e t of msulation r ical Specifications requirements on operating limits during heatup, cooldown, and E .T I
$ system hydrostatic pressure testing shall be employed for these tests.
h5 5.2.4.7 Code Exemptions As provided in ASME Section XI, IWB-1220, certain portions of Class 1 systems are exempt from the volumetric and surface examination requirements of IWB-2500. The followmg components (or parts of components) are exempt from the volumetric and surface examination requirements ofIWB-2500:
- 1. Components that are connected to the reactor coolant system and part of the reactor coolant
, pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the reactor coolant system under nornul plant operating conditions is within the capacity of makeup systems which are operable from on-site emergency power; e r p'png of 1 inch nominal pipe size and smaller, except for steam generator tubing; o components and their connections in piping of 1 inch nominal pipe size and smaller;
- 2. reactor vessel head connections and associated piping,2 inch nominal pipe size and smaller, made inaccessible by control rod drive penetrations.
5.2.5 Reactor Coolant Prissure Boundary IAnkage Detection Systems Means for the detection of leakage from the Reactor Coolant Pressure Boundary are provided to alert operators to the existence ofleakage above acceptable limits, which may indicate an unsafe condition for the facility. 'Ibe leakage detection systems are sufficiently diverse and sensitive to meet the criteria of A Regulatory Guide 1.45 for leaks from identified and unidentified sources. 'Ihe leakage detection systems are capable of performing their functions following seismic events that do not require plant shutdown. Ammed cemen namenw acs and conneend am noe s.s-as
System 804- DCD - Potential Design Changes i Item Number: 2 Summary Descrintinn: NPSH for SCS and CSS Pumps Affected DCD Sections: CDM: None ADM: Section 6.5.3.4 Figures 6.3.2-1 A and 6.3.2-1B Descrintion of Chanvei Reason for Change: There are two root causes for this design change:
- 1. While bidding System 80+ for the Lungmen Project in Taiwan, subsequent to the issuance of the Final Design Approval, ABB-CE received data from prospective pump vendors KSB and )
Ingersoll-Dresser. The data showed that the net positive suction head required (NPSHR) by the proposed shutdown cooling (SC) and containment spray (CS) pumps exceeded the NPSHR assumed in preparation of DCD Section 6.5.3.4. We had assumed a maximum NPSHR of 20 feet at pump runout flow of 6500 gpm. The vendors require approximately 25 feet at 6500 gpm.
- 2. During the engineering work to support CESSAR-DC, ABB-CE did not evaluate the net positive suction head (NPSH) available to the SC pumps when aligned for containment spray.
As a result, two design concerns were identified:
- 1. There may be insufficient NPSH to the containment spray pumps if they operate at the ,
maximum flow rate of 6500 gpm stated in the DCD. l
- 2. There may be insufficient NPSH to the SC pumps when aligned for containment spray, due to large frictional losses caused by the length and diameter of the suction crossover piping.
[The NPSH available to the SC and CS pumps during all other operating modes is adequate.]
Description:
The NPSH during containment spray operation was evaluated and some SCS/ CSS
. suction lines were re-sized so that the minimum available NPSH exceeds the NPSHR specified by the pump vendors. Based on the revised NPSH calculations, and as shown on the attached markup of DCD Figures 6.3.2-1 A and IB, the changes are:
. Increased the nominal diameter.of CS suction piping, including valves SI-104,105,157, and 158, from 18 to 20 inches e lacreased the nominal diameter of CS/SC pump suction crossover piping, including valves SI-340 and 342, from 18 to 20 inches e Increased the nominal diameter of SC suction piping, including valves SI-107 and 106, from 14 (or 18) to 20 inches Despite increasing these line sizes, it was not possible to obtain enough NPSHA for the CS or SC pumps during containment spray operation above 6000 gpm. Therefore, the maximum allowable TKSdedch. doc 2 6/11/96
System 80+ DCD - Potential Design Changes t l .
- containment spray flow rate was reduced from 6500 gpm to 5500 gpm. Safety and containment l analyses use CS flow rates of 5000 gpm minimum,6500 gpm maximum, so this change provides margin to the maximum assumed value.
l During the performance of these analyses, it was also determined that the IRWST water volume presented in DCD Section 6.5.3.4 was the volume above the pump suction piping. The volume will be represented as to the total volume remaining in the IRWST (198,000 gallons), to be consistent with the measurement in Technical Specification 3.5.4. The minimum IRWST level (75.5 feet) is unchanged. ! l The following table compares original and revised NPSH values for containment spray l operation: l NPSH Required (Feet) Minimum NPSH Available (Feet) Flow Rate Assumed in Vendor CS Pump CS Pump SC Pump SC Pump (gpm) DCD Data Original Revised Original Revised 5000 none 16.5 24 21.9 n/c 20.7 l 5500 none 18 n/c 21.1 n/c 19.6 l 6000 none 20 n/c 20.3 n/c 18.5 6500 < 20 25 21.2 19.4 n/c 17.4 1 There are no design changes to the pumps themselves. The SC and CS pumps are still identical and interchangeable. The minimum pump design head and design flow will be confirmed by ITAAC, as presented in Certified Design Material Tables 2.3.2-1 and 2.4.6-1. This change has no impact on safety. In fact, it refimes the System 80+ design to ensure that the ! CS and SC pumps are more likely to meet their safety functions. DCD Markuns Attached? Yes l TKSdedch. doc 3 6/11/96 l l t
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Syst m 80+ Re9 Cis o K & . :r'- Design ControlDocument b . GM Figure 6.2.1-16), the long-term airborne elemental conientration is calculated to be less than 20 percent
. f 3 e:pf the organic. Thus, depression of the sump pH due to the long-term production of hcl by the Q p o irradiation of the electrical insulation and the radiation-induced nitric acid formation would not impact jw g the dose assessment.
s 4 j De transient spray removal lambdas for the 10 CFR 100 LOCA analysis are shown on Figure 6.5-6. 6.5.3.4 Available Net Positive Suction Head (NPSH) h y The IRWST is the suction source for the SI pumps and CS pumps during short term injection and long tr 4 term cooling modes of post-accident operation. As described in Section 6.8, the Holdup Volume Tank
% (HVT) performs water collection services after an accident. Spillways allow accumulated water in the Q HVT to spill back into the IRWST, thereby replenishing IRWST water volume during accident
, j operations. The minimum available NPSH for the SI and CS pumps was determined based on the e minimum water level in the IRWST during accident conditions. In addition, the following conservative
@ massumptions are made:
%. R
{g, go Fluid conditions in the IRWST are saturated; no credit is taken for an increase in containment pressure. 1e The contribution of the volume of water spillage from the RCS and safety injection tanks is l
, f.
i conservatively neglected. l qhe With the CS system actuated, the reactor cavity is assumed flooded and the HVT full to a level R $,g that is just below the level at which water begins to return to the IRWST through the spiilways.
- Spray water is being held up on surfaces throughout *he containment. Locations for the accumulation of water inside the containment includ: water held up on horizontal surfaces, clogged floor drains, water held up in containment spray piping, water in the containment atmosphere, water film on vertical surfaces, puddles trapped on equipment, water soaked into insulation, and the contamment free volume filled with steam.
The SI and CS pumps are located in the reactor building subsphere and are placed low enough below the minimum IRWST fluid level to assure adequate available NPSH. The minimum IRWST fluid volume aftar
- Mdent has been determined to be , gallons. His corresponds to a water level elevation of]&dfeet. y'a 74 5~ j.3 ne~cateuTated available NPSH for the s ranges fr J4' et at the design flow rate of 5000 gpm 3d[ t .. f t at a pump runout flow 4MKTgp . This excee CS pump required NPSH ofKfeet
'"**y;i9 He calcufted available NPSH for the SI ps is fa 84.e-9 p 26.9 feet at a pump runout flow of 1235 gpm. This exceedsJNe SI pum r *red NPSH of'20 feet at runout flow.
8rM8d , 74.fr During a LOC ' tie e or cavity will not be flooded as was assumed in determining the minimum water level of " feet. Because of this, an additional volume of water will be available to raise the minimum water level in the IRWST by approximately 2 feet, thereby increasing the available NPSH. Prevention of the entrance of debris into the IRWST and SI and CS pump suction lines is described in Section 6.8. l l Anwoved Design Material Engineered Safety Features Pope 6.623
4 System 80+ DCD - Potential Design Changes Item Number: 3 Summary
Description:
Revise Safety and Seismic Classification of SDS Components l from PSVs to IRWST ' Affected DCD Sections: CDM: None ADM: Tables 3.2-1 and 3.2-2 Description of Channe: l 1 i This is a consistency change to correct errors, not a design change. Reason for Change: The safety and seismic classification of the SDS spargers, vacuum breakers and piping from the SDS valves and PSVs to the IRWST were specified inconsistently. ABB-CE desires to change classifications so that all components in this portion of the SDS are classified consistently.
Description:
This change revises the safety class and seismic category of the spargers and vacuum breakers, and upgrades the seismic category of the piping, so that they have the same safety classification and seismic category. For the spargers and vacuum breakers, the safety class will change from 2 to Non-Nuclear Safety (NNS) and the seismic category will change from I to II. For the piping, the seismic category will change from non-seismic (NS) to II. The quality I class will change from 1 to 2 for a!! NNS components. This change has no impact on safety: The spargers are not pressure retaining components, and i do not perform and safety-related function. Changing the safety classification from 2 to NNS changes the design code from ASME B&PV Section III to ASME/ ANSI B31.1, which has similar design rules for structural integrity. Changing the seismic category from I to II requires i that the spargers still maintain sufficient integrity during seismic events, such that they would not ! damage safety-related equipment during the Safe Shutdown earthquake. l The vacuum breakers do not perform a safety-related function. Changing the seismic category from I to II requires that the vacuum breakers still maintain sufficient integrity during seismic events, such that they do not damage safety-related e'quipment. This change provides consistent safety, seismic, and quality classification for these components. As currently specified, there would be several interfaces between safety and non-safety equipment, which would be complex to design and construct. The proposed revisions would provide consistent interfaces and higher assurance of correct construction. l i DCD Markups Attached? Yes l TKSdedch. doc 4 6/11/96
% Mo.] . h hqp g.
System 80+ Ossian controlDocument Table 3.2-1 Classification of Structures, Systems and Cotnponents O ~ Safety Seismic Quality l f%npananr Identincadon Class Category I4 cation
- Chn*
Ranctor Coolant Systen l Reactor Vessel ! I RC 1 Stamm Gensemors (prunary/szondary) 1/2 (1]* 1 RC 1 ] I RC l Presseruer 1 1 a l Reactor Coolant Pumps (2,3,9]* 1 I RC 1 ;
- Pipeg within Ranctor Coolartt Pressure 1/2 [4] I RC 1 1 Boundary (5]
! Control R1===t Drive Macb=- [6] [6] RC 1 , l Core Suppoit Structures and 3 I RC 1 l 1 j Internals Senusares (7] Fuel Aassablies (8] 2 I RC 1 Control Element Assemblies [8] 3 I RC 1 RC Closure Head LiR Rig NNS II(10] 2 l Heated Junction Thermocouple Probe 1/3(1;I I RC 1 i Asundh l HTIt Pressure Houang 1 ! RC 1
- ICI Cable Trsy Support Frecas 3 I RC 1 l ICI Holding Frame NNS NS RC 3
- ICI Quids Tubes 1 I RC 1
. ICI Guide Tube Supports ICI Sent Housmg 1 1 I I RC RC 1 1 m i 1
- ICI Seal Table 1 I RC 1
- Piping (27] 1/2 I RC 1 Valves (27] 1/2 I RC 1 i Is>contelament Water Storage Systen l IRW8T 3 I RC 1
' Holdop Volume Tank 3 I RC 1 Presame Relief Denpers 3 I RC 1 Cavity Flooding System j Piping 2 I RC 1 i Valves 2 I RC 1
- Safdy mE d d- Systen j
Valves 1/2/M ' I/K RC t% 1[15 2. i
- aping 1/2/NNS IAINT.- RC
] Spargers fNNS /E RC 73 l i Safety Inqlectica Syster 'a } Safety Inqledios Pumps 2 I RB 1 j Safety Iedaction Tanks 2 I RC 1 l Piping [24,27] 1/2 I RBIRC 1 {. Valves [27] 1/2 I RB/RC 1 0 Refer to Notes at and of table. Wesage Assesaw. ass 4m erasc tamst Aue AN
I System 80+ Desian controlDocument O V Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality Component Identification Class Category Location" Class" W Station Air System Air Compressors NNS NS SB 3 Air Dryers / Filters NNS NS SB 3 Air Receivers Piping [27] NNS 2/NNS NS I/NS SB All 3 1/3 k Valves [271 2/NNS UNS All 1/3 Breathing Air System Air Compressors NNS NS SB 3 Piping [27] 2/NNS UNS All 1/3 Valves [27] 2/NNS UNS All 1/3 Air Receivers NNS NS SB 3 Air Dryer / Filters NNS NS SB 3 Compressed Gas Systems High Pressure Gas cylinders NNS NS YA 3 Pressure Regulators NNS NS YA 3 Imak Detection Systems NNS NS All 3 l Liquid Nitrogen Evaporators NNS NS YA 3 i p) Piping [26,27] Valves [27] 2/NNS 2/NNS UNS I/NS All All 1/3 1/3 Fire Protection System ; Jockey Pump NNS NS FP 2 I Backup Storage Tank NNS I NA 1 ! Fire Pumps NNS NS FP 2 l Backup Fire Pump NNS I NA 1 l Storage Tanks NNS NS FB 2 l Water Spray Systems (Deluge and 2/NNS 1/II/NS TB/NA/RC/RB/ 1/2 l Sprinkler) Piping, Valves [16, 27] DG/SB p ! Hose Systems / Standpipes [16,27] 2/NNS UNS All 1/2 , Portable Fire Extinguishers [16] NNS NS All 2 l Exterior Distribution System Piping NNS NS YA 2' Valves NNS NS YA 2
't Stramers NNS NS YA 2 Alternate AC Sourte/ Combustion NNS NS YA 2 Turbine-Geneestor
.cj DG Engme Maud Oil System [17]
Fuel Oil Storage Tanks 3 I DF 1 I ( Recirculation Pumps NNS NS DF 3 Booster Pumps 3 I DG 1 ! Fuel Oil Day Tanks 3 I DG 1 emegeug M$c\ Cimenb 5%w) y
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System 80 + Design contro/ Document Table 3.2-1 Classification of Structures, Systems, and Components (Cont'd.) Safety Seismic Quality p-l Component Identification Class Category Location
- ClassM Main Steam Supply System Piping [21]
Steam Generator to MSIV's 2 I RC/MS 1 Other NNS NS MS/NA/TB 3
- 1. d Main Steam Supply Systan Valves [21]
Safety Valves 2 I MS 1 MS!V's, MSIV Bypass Valves 2 I MS 1 Atmospheric Dump Valves 2 I MS 1 Valves 2/NNS UNS NA/MS/TB 1/3 Containment flydmgen Recombiner System Hydrogen Recombiners 2 1 NA 1 Hydrogen Analyzers 2 I NA 1 Hydrogen Recombiner Control Panel 3 I NA 1 gl Piping [27] 2 I NA/RC 1 Valves [27] 2 I NA/RC 1 Steam Generator Blowdown System [22] Flash Tank NNS NS TB 2 Heat Exchanger NNS NS TB 2 Filter NNS NS TB 2 Demineralizers NNS NS TB 2 u-l Piping [27] 2/NNS I/NS RC/TB/MS 1/2 ! Valves [27] 2/NSS I/NS RC/TB/MS 1/2 : Steam Generator Wet Layup ! Recirculation Systan [22] l Piping [27] 2/NNS 1/NS RC/TB/MS 1/3 I Valves [27] 2/NSS I/NS RC/TB/MS 1/3 l Hydrogen Mitigation System Hydrogen Igniters NNS I RC 2 ( Bdt> > Potable and Sanitary Water Systems NNS NS YA 3 f 1 Instrumentation and Contml Systems j Plant Protection System (PPS) y The PPS includes the electrical and mechanical devices and circuitry (from sensors to actuation device input termmals) involved in generating the signals associated ; with the two protective functions < defined below: Approved Des &n Matenal- Design of SSC -42M61 Page 3.2-18 P/%)
I System 80+ Desian ControlDocument f~ Table 3.2-1 Classification of Stnictures, Systems, and Components (Cont'd.) Safety Seismic Quality Component Identification Class Category Location" Class" l" Nuclear Annex Structure Control Area 3 I NA 1 l-EFW Tank / Main Steam Valve House 3 I NA 1 Area Emergency Diesel Generator Areas 3 I NA 1 CVCS/ Maintenance Area 3 I NA 1 Fuel Handling Area 3 ,P I NA 1 ( Other Simctures NNS II NA/RB 2-l" Unit Vent Turbine Building NNS II TB 2 Radwaste Building [28] NNS #1 RW 2 l Station Service Water Pump / Intake 3 I SP 1 Structure Component Cooling Water Heat 3 I CX/YD 1 , Exchanger Structures and Pipe Tunnels Diesel Fuel Storage Structure 3 1 DF 1 Station Services Building / Auxiliary NNS NS SB 3 boiler Structure Administration Building NNS NS ADB 3 Warehouse NNS NS WH 3
- 0' Fire Pump House NNS NS FP 3 Alternate AC Source / Combustion NNS NS YA 2 Turbine 4enerator Structure and Fuel Tank Dikes Dike (Holdup. Boric Acid Storage and NNS II YA 2
- Reactor Makeup Water Tanks) [28]
Dike (Condenaste Storage Tank) [28] NNS II YA 2
- t. Cranes
! Polar Crane NNS II RC 2 l Cank Handling Hoist NNS II NA 2 l
-' New Fuel Handhng Hoist NNS II NA 2 j Component Supposts [23] 1/2/3/NNS 1/NS All 1/2/3 l-< ,
\
i 4 t G./%) Annmed Dadgn A0etend - Denign of Ssc waEl Page 3.2 21
1 l Oha' & hsev /^b Svstem 80+ Desian controloccument 1 Table 3.2-2 Safety Class 1,2 & 3 Valves (Cont'd.) i I l Component Location / Safety Sdenic Quality I hification Description Class Category Class aNot Coollag and Purification Systern (PCPS) [1] (Cont'd.) PC-M8, 209 Coohng pump discharge isolation 3 I 1 PC-211,212 Cooling HX inlet isolation 3 I 1 PC-213, 214 Cooling HX outlet isolation 3 I 1 PC-249 IRWST reurn line isolation 3 I 1 PC-257, 258 Refueling pool discharge isolation 2 1 1 PC 291,292 Refueling pool inlet isolation 2 1 1 PC-300, 301, 302, 303 Cooling flow indication isolation 3 I 1 PC-320, 321 Cooling puinp suction pressure 3 I 1 Safety Deprsusurization Systen (SDS) RC-406, 407, 408, 409 Rapid depressuruation 1 I 1 RC-410, 411, 412, 413 Pressuruer vent 1 i 1 RC-414, 415, 416, 417 Reactor vessel vent 1 1 1 RC-418 RCGVS vent to RDT 2 1 1 O RC-419 RCGVS vent to IRWST 2 1 1 RC-263, 264 RD pressure inardion 2 1 1 RC 267 RCGVS pressure indication 2 I 1 A hC-XXX s \ MS/ Safety ' valve sparger IErie'l \ 2' h - QdP \ 'N vacuum breaker'% s \ ON d . % 7ch^ um N N '
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SI-104,105 CS pump suction isolation 2 1 1 SI-106,107 SCS pump sudion isolation 2 I 1 SI-108,109 SCS pump suction pressure indication 2 1 1 isolation SI-113, 123, 133, 143 Safety injection containment check 2 I 1 SI 115, 125, 135, 145 SI flow indication isolation 2 1 1 SI-116, 126, 136, 146 SI flow indirmion isolation 2 I 1 SI-117, 127, 137, 147 SIT pressure indweian isolation 2 I 1 p SI-119, 129, 139, 149 S1T pressure indication isolation 2 I 1 Anorecentamen nanamnar oneen er sac __ _ hoe a.2 32
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System 80+ DCD - Potential Design Changes Item Number: / S" Summary
Description:
Additional CEA locations in Reactor Core Affected DCD Sections: CDM: Yes, Figure 2.2.1.3; eliminate the maximum number of CEDMs in Section 2.2.2 ADM: Yes, Figure 4.2.11, Figure 4.3-46, Figure 4.3-47, Text in Section 4.2.2.4 Descrintion of Chance: The objective of the change is twofold: (1) allow for the possibility of having 4 element CEAs at twelve specific core locations; and (2) allow for the possibility of replacing 4-element CEAs with 12-element CEAs at for specific core locations. Part (1) of the change is accomplished by identifying twelve core locations (in addition to the eight core locations previously identified) for spare locations which may contain 12 element CEAs instead of 4 element CEAs. Not all of the locations identified for spare 4 element CEAs would necessarily contain CEAs. The minimum number of CEAs would remain at 93. The proposed change would allow flexibility in choosing locations for additional CEAs and/or CEA elements. The addition of CEAs in the central core region would permit enhanced maneuvering capability. The addition of CEAs or CEA elements could also improve shutdown margin. All locations identified for additional CEAs satisfy the minimum reactor vessel head ligament requirement. The ADM (Section 4.1.1) currently allows changes to certain features and evaluated parameters for the fuel system design, nuclear design, and thermal and hydraulic design of the initial core without prior NRC review and approval provided these changes are within certain acceptance criteria. With the propose change, Section 4.1.1 would remain unchanged. The CDM (Figure 2.3.1-3) and the ADM (Figure 5.3-7) show a minimum of 103 CEDM and instrumentation nozzles, and a minimum of 2 HJTC probes. The proposed change would maintain these minimum values. Since specific CEDM locations are not shown in the ADM, the ,
~
change permits the selection of 8 different core loc ~ations for spare CEAs from those currently j shown, as well as the addition of spare CEAs at other core locations. ! i
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DCD Markuns Attached? Yes 1 1 l I TKSdedch. doc 5 6/11/96 3 I
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System 80+ Desian controlDocument 2.2.2 Control Element Drive Mechanism Design Description De control element drive mechanism is a magnetic jack device that positions and holds the control element assemblies relative to the fuel assemblies. The primary safety-related function of the Control Element Drive Mechanism (CEDM) is to release the Control Element Assembly (CEA) upon termination of electrical power to the CEDM. A minimum of 93 CEDMs is required, hc =:, 2 mW~"- M = h"-td c..c CEOM cou bc umolied, ne CEDM also acts as a primary pressure boundary as part of the Reactor Coolant System. Refer to Section 2.3.1 for CEDM primary pressure boundary aspects. Inspections, Tests, Analyses, and Acceptance Criteria None The initial test program addressed in Section 2.11 will test the ability of the CEDM to release the CEA upon termination of electrical power to the CEDM. The Basic Configuration of the CEDM primary pressure boundary components will be verified as part of Section 2.3.1. The CEDM pattern will be verified as part of Section 2.2.1. I I
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d ( 1 System 80 + and/c '.EA weh oesian control Documen-i L a spider structure which couples to the contro tomom ho r ism (CEDM) drive shaft extension. He neutron absorber elements of a four-element CEA engage the four co rqide tubes in a single fue ' assembly. He four-element CEAs are used for control of power distribution and core reactivity in the power operarmg range. The twelve-element CEAs engage the four corner gt}ide tubes in one fue , I assembly and the two nearest corner guide tubes in adjacent fuel assemblies. - The twelve-element CEA! ' make up the balance of the control groups and provide the core with strong shutdown rods. He contro element assemblies are shown in Figures 4.2 3 through 4.2 5 and Figure 4.2-F4. The pattern of CEA: (total of 9.;) is shown in Figure 4.211. F _ _ y m Miditional CEAs Unay be installed if desirec ' for additional flexibility or future use. Twenty-five of the 93 CEAs are part-strength CEAs (PSCEAs) l Part-strength CEAs are differentiated from full strength CEAs by using alphanumeric serialization insteac of the numerical system used on the full-strength CEAs. l All control elements are sealed by welds which join the CEA cladding to an Inconel 625 nose cap at tht i bottom, and an inconel 625 connector at the top which makes up part of the end fitting. De end fittings in turn, are threaded and crimped in place by a locking nut to the spider
- lateral and axial support for the control elements. The spider hub bore is specially machined to providi <
a point of attachment for the CEA extension shaft. He control elements of a twelve-element full-strength CEA consist of an Inconel 625 tube loaded wit! l a stack of cylindrical absorber pellets. The absorber material consists of 73% TD boron carbide (B C l pellets, with the exception of the lower portion of the elements, which contain reduced diameter Bf r pellets wrapped in a sleeve of Type 347 stainless steel (felt metal). He design objective realized by the use of felt metal and reduced diameter B C 4 pellets in the elemen tip zones is that as the B 4C pellets swell due to irradiation, the felt metal sleeve compresses as a resul of the applied loading. His compression limits the amount ofinduced strain in the cladding. Derefore buffering of the CEA following scram, which occurs when the element tips enter the reduced diamete: l portion of the fuel assembly guide tubes, is not affected with long term exposure of the CEA to reacto j operating conditions. l I During normal power operation, all of the twelve-element CEAs are expected to be in the fully withdrawr l position. Rus, the local B-10 burnup progresses at a lower rate, and CEA life is prolonged. Above th: l l absorber column is a plenum which provides expansion volume for helium released from the B.C. nc , l ' plenum volume contains a Type 302 stainless steel holddown spring, which restrams the absorber matena agamst longitudinal shifting with respect to the clad while allowing for differential expansion between tit ,
, absorber and the clad. He spring develops a load sufficient to maintain the position of the absorbe. ;
material during shipping and handling. - De control elements of a four-element full-strength CEA consist of an Inconel 625 tube loaded with : stack of cylindrical Ag-Intd absorber bars. This CEA design is used for the regh! banks. Twi design objectives are realized by use of Ag-In-Cd absorber over the full active length: e CEA Cladding Dimensional Stability } l Because of its high ductility and low strength, the Ag-In-Cd will not deform the CEA cladding l ' Buffering of the CEA following scram, which occurs when the corner elaamar tips enter the j reduced diameter portion of the fuel assembly guide tubes, is not degraded with long-tern i exposure of the CEA to reactor operating conditions. i Am==W osava messame nosesor esos u -t
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i l System 80+ DCD - Potential Design Changes Item Number: [h l Summary Descrintion: Add RCS Mid-loop Level Measurement Tank and Instrumentation Across the Hot Leg l Affected DCD Sections: CDM: None ADM: Chapter 5 Chapter 7 Appendix 19.8A Description of Chance: A detailed engineering evaluation was performed to confirm the feasibility of adding tanks connected directly across each of the RCS hot leg pipes to measure the hot leg water level during Mid-Loop operations using IlJTC instruments. The current DCD Chapter 5 includes the RCS P&ID and the cross section of the Reactor Vessel. The Shutdown Risk evaluation reported in DCD Chapter 19.8A provides a figure (Figure 2.8-4) which indicates that two additional sets of liJTC's are inserted through the reactor vessel head to measure the hot leg water level. The DCD text indicates this level should be measured to an accuracy of plus or minus 1 inch. In Chapter 19 the tank across the hot leg will be shown for measuring the hot leg water level during mid-loop operations. The water level will be measured with heated junction thermocouples that i I jerate in the same manner as the ICCMS HJTC located in the reactor vessel. The reasons of the proposed change are as follows:
- 1. the instruments are permanently installed and are not affected by refueling activities in the reactor vessel head area.
l
- 2. the reliability of the instruments is enhanced by relocating them to a more benign area.
- 3. there is improved flexibility for plant operations and maintenance for servicing the j instruments and installing nozzle dams.
- 4. the water level measurements are taken closer to the Shutdown Cooling suction line ~
which is the point ofinterest. i Locating the Mid-Loop HJTCs through the rer.ctor vessel head have limitations that the proposed change eliminates. The cable to the head mounted instruments must be disconnected when the reactor vessel head is removed. The disconnection affects the ability of the instruments to be used. Having the instruments pennanently connected reduces the potential for human error. l Since the position of the reactor vessel head (and components attached to it) changes when the head is detensioned, the instruments should only be used with the head tensioned. The amount i of position change is a function of the fuel burnup and is therefore not constant from fuel cycle to fuel cycle or from before refueling to after refueling. Relocating the instruments will reduce human error and increase operational flexibility. The current design reduces the flexibility in TKSdedch. doc 6 6/11/96
l l - ! System 80+ DCD - Potential Design Changes l l l when the nozzle dams can be installed or removed from the steam generator. The proposed change allows more flexibility and will result in the nozzle dams being installed for a shorter period. The location of the instruments in the reactor vessel means that they are exposed to reactor l pressure, temperature and radiation. These conditions will decrease the operating life of the instruments compared to locating the instrument in the tanks. Instrument replacement in the reactor vessel is more difficult and results in higher personnel exposure. The Mid-Loop HJTC System uses the proven Heated Junction Thermocouples to measure the water level that are the same design as the ICCMS HJTCs. The HJTC heater controllers are the same as the ICCMS HJTC heater controllers and each controller is connected to a separate power supply. Placing the HJTC instruments in the tank allows a larger instrument diameter which results in higher resolution of the HJTC by using radial distance to accomplish thermal separation of the HJTCs. The instrument operating environment is at atmospheric pressure and ! temperatures of less than 212 F which will increase the operating life and increase reliability. j The water level is indicated in the Control Room. l There is a fluid connection from the bottom of the hot leg to the bottom of the tank. An air
- connection from the top of the hot leg to the top of the tank causes an equalization of the water I level in the hot leg and tank during mid-loop conditions. There are two (2) isolation valves on l each of the connecting lines. The connecting lines up to and including the second isolation valve l are designed RCS conditions using the ASME B & PV Code, Section III, Class 1. The remaining portions of the connecting lines and the tank up to and including the tank drain valve
. are designed for reactor operating temperature and pressure in accordance with ASME B & PV Code, Section VIII. The drain line from the tank connects to the liquid radwaste system.
i DCD Markuns Attached? Yes l l \ i i 5 TKSdedch. doc 7 6/11/96 l ,. _ , . _ _ _ _ _ , -
System 80+ Design controlDocument 5.0 Reactor Coolant System and Connected Systems 5.1 Summary Description The reactor is a pressurized water reactor (PWR) with two coolant loops. The reactor coolant system (RCS) circulates water in a closed cycle, removing heat from the reactor core and internals and transferring it to a secondary system. The steam generators provide the interface between the reactor coolant (primary) system and the main steam (secondary) system. The steam generators are vertical U-tube heat exchangers with an integral economizer in which heat is transferred from the reactor coolant to the main steam system. Reactor coolant is prevented from mixing with the secondary steam by the steam generator tubes and the steam generator tube sheet. making the RCS a closed system thus forming a barrier to the release of radioactive materials from the core of the reactor to the secondary system and containment building. The arrangement of the RCS is shown in Figures 5.1.31 and 5.1.3-2. The major components of the system are the reactor vessel; two parallel heat transfer loops, each containing one steam generator and two reactor coolant pumps; a pressurizer connected to one of the reactor vessel hot legs; and associated piping. All components are located inside the containment building. Table 5.1.1-1 shows the principal pressures, temperatures, and design minimum flowrates of the RCS under normal steady-state, full-power operating conditions. Instrumentation provided for operation and , control of the system is described in Chapter 7. l System pressure is controlled by the pressurizer, where steam and water are maintained in thermal equilibrium. Steam is formed by energizing irnmersion heaters in the pressurizer, or is condensed by the pressurizer spray to limit pressure variations caused by contraction or expansion of the reactor coolant. The average temperature of the reactor coolant varies with power level and the fluid expands or contracts, changing the pressurizer water level. The charging pumps and letdown control valves in the chemical and volume control system (CVCS) are used to maintain a programmed pressurizer water level. A continuous but variable letdown purification flow is maintained to keep the RCS chemistry within prescribed limits. A charging nozzle and a letdown nozzle are provided on the reactor coolant piping for this operation. The charging flow is also used to alter the boron concentration or correct the chemical content of the reactor coolant. Other reactor coolant system penetrations are the pressurizer surge line in one hot leg;pe four direct
. vessel injection nozzles in the reactor vessel for the safety injection system; two return nozzles to the shutdown cooling system, one in each hot leg; two pressurizergray '"# ** *nozzles;
- vent and drain connectio and sample and instrument connections, /uM9 hw Overpressure protection for the reactor coolant pressure boundary is provided by four spring-loaded ASME Code safety valves connected to the top of the pressurizer. These valves discharge to the in-containment refueling water storage tank, where the steam is released under water to be condensed and cooled. If the steam discharge exceeds the capacity of the in-containment refueling water storage tank, it is vented to the containment atmosphere.
Overpressure protection for the secendary side of the steam generators is provided by spring-loaded ASME Code safety valves located in the main steam system upstream of the steam line isolation valves. Approved Design Material RCS and Connected Systems Page 6.1 1
System 80+ Deslan ControlDocument range DP sensors measure RCS level in the hot leg region. He narrow range instrumentation includes low and low-low alarms which annunciate in the control room. He wide range DP sensors measure RCS level in the hot leg to the top of the pressurizer. He wide range instrumentation also includes low and low-low alarms which annunciate in the control room. He indication and a?. arms allow the operator to monitor RCS level from the control room during shutdown operations which require reduced RCS
$ inventory, (s. Two redundant Refueling Water Level Probes (RWLP) provide independent level indic N: of the vessel to the fuel alignment plate. The narrow range RWLP assemblies measure reactor coolan liquid inventory in the hot leg region during reduced inventory periods when the reactor head is installed.
N De wide range RWLP assemblies measure reactor coolant liquid inventory in the upper pcrtion of the 4 vessel to the fuel alignment plate. He basic principle of operation is the detection of a temperature difference between one of the heated thermocouples and the unheated thermocouple at the bottom of the k RWLP assembly. Each RWLP assembly includes multiple HJTC sensors, r.n outer sheath, a seal plug a and electrical connectors. A -=r. 3 De RWLP thermal hydraulic operating environment is relatively uncomplicated. The narrow range M RWLP is used only during non-power operation while the reactor vessel head is in place. De probe assembly is housed in a stainless steel structure that protects it from flow loads. De narrow range RWLP heated junction thermocouple sensors are more closely spaced in the hot leg region to provide improved resolution. The RWLP's provide indication, high, high-high, low, and low-low alarms in the contml room. B. l RCS temperature is measured using the existing CET temperatures, HJTC unheated sensor temperatures, and RCS Hot Leg RTD temperatures. De CETs and HJTC unheated sensors have high and high-high alarms to annunciate the approach to bulk boiling in the core. De RTDs have a high alarm anmmeintian. De HJTC unhastad sensor temperature is not available when the head is off. Each train of the SCS has a measurement of SCS flow. Dis measurement pmvides indication of return flow to the RCS when either the SCS pump or CS pump is being used for shutdown cooling. Low flow is annunciated in the control room. To monitor the p fessance of the SCS and CS pumps, pump suction pressure, discharge pressure and motor current are monitored C. annunciated in the control room. The performance of the SCS heat exchaqer is monitored and annunciated by measuring the issp .ute
'. in the inlet and murn lines. Valve pwtlon indication is provides indication of the system lineup and provides the status of the available flowpaths.
7.7.1.1.16 Steam Generator Tube Rupture Detection Instrumentation System 80+ incorporates N-16 gamma detection with a scintillation detector and microprocessor based signal conditioning on a header leaving each steam generator. A description and the applicability of using N-16 gamman in de'=ing steam Generator tube leaks is provided in Section 5.6 of Appendix 5F. De datanion system will alert the operator to a SG tube leak condition originating at power and identify which ste in generator is affected. The addition of N-16 radiation detection and monitoring equipment further enhances the diagnosis of steam generator tube leaks or ruptures and provides the operator with more accurate information to assess Anomed Demon anneenet - haneannentesen ear conow n o ,1.1.so
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+
l ( 1 . System 80+ DCD - Potential Design Changes Insert in 7.7.1.1.15 The first HJTC system displays the output from the two'(2) inadequate core cooling probes located inside the reactor vessel that are availabe when the reactor vessel head is installed. The inadequate core cooling probes are described in Section 7.5.1.1.7. The second HJTC system measures the water level in the RCS hot leg pipes during Mode 5 reduced inventory conditions. This Mid-Loop HJTC system consists of an instrument installed , in a tank attached to the RCS hot leg pipe. There is a separate tank and instrument attached to l each hot leg in the vicinity of the SCS suction connection. There are two (2) connections each l tank and corresponding hot leg. One connection is at the bottom of the hot leg to the bottom of ; l the tank. The second connection is at the top of the hot leg to the top of the tank. When the isolation valves on the connections are open, the water level in the tank equalizes with the water level in the tank. The isolation valves are operated from the control room. The position of the
- valves is indicated in the control room.
The connecting pipe, up to and including the second isolation valve, is designed to the same conditions as the Reactor Coolant System and is ASME B & PV Code, Section III, Class 1. The tank and connecting pipe after the second isolation valve is designed for RCS operating pressure and temperature in accordance with ASME B & PC Code, Section VIII up to and including the ! tank drain valve. l During Mode 5 reduced inventory conditions, the Mid-Loop HJTC system is connected to the RCS by opening the isolation valves. The RCS mid-loop vent path described in Chapter 19.8A, 4 Section 2.3.3.3 assures the system pressure at near to atmospheric and the fluid temperature at less than 212 F with no boiling. The Mid-Loop HJTC instrument consists of a vertical array of heated junction and unheated junction thermocouples. The heated junction thermocouples are spaced to obtain the required i measurement resolution and provide alarm points for high level (water level approaching the ! steam generator nozzles) and low level (water level approaching loss of SCS suction). The l thermocouple design is described in Section 7.5.1.1.71.2. Since there is no two phase conditions j in Mode 5 at the tank, the instrument does not include the phase separator tubes included in the l ICCMS HJTC. The are separate heater controllers for each Mid-Loop HJTC instrument. Each
; contrclier is connected to a separate power supply to mitigate conunon mode failure.
l The water level in the RCS hot legs is displayed in the control room. TKSdedch. doc 11 6/11/96 l _ _
System 80+ Design ControlDocument e mitigation p'anning aimed at the reinitiation of shutdown cooling, delaying the onset of boiling, and delaying core uncovery. The design goals of the instrumentation package are to provide: Prevention - enhanced monitoring capabilities for prevention of a complete loss of SCS operation, and Mitig: tion - the timely response to a loss of SCS. These goals have been achieved with the design features of the System 80+ instrumentation described in the following. 2.8.3.2 Instrumentation Description kl 4 ' Table 2.8-1 describes the instrumentation package for reduced inventory operations included in the System 80+ design. Additional details are provided below. i ! 2.8.3.2.1 12 vel S}h
%g il Four unique sets of instruments are provided for the measurement of level during RCS draindown and
- 43 reduced inventory operations. These instruments make up the retueling water level indication system, i]
14 The first set of instruments is a pair of wide-range, dP-based level sensors. These sensors are provided g to measure level between the pressurizer and the junction of each SCS suction line 'vith the RCS during e draindown operations. Another pair of dP-based level sensors is utilized to determine RCS water level gI; once it is within the reactor vessel. These narrow-range level sensors function to measure level between 9 the direct vessel injection (DVI) nozzle and the junction of the SCS suction lines with the RCS. y, Fit 2 l One wide-range and one narrow-range dP instrument are connected to each SCS suction line. Separate lower level taps are provided for each instrument. See Figure 2.8-2. Because of the location of the 4* *k ! upper level taps, each of these dP instrur.'ents will operate with, or without, the reactor vessel head in place, f3g t In addition to the dP-based instruments described above, two heated-junction thermocouple (HJTC) q} systems O also be available fo reactor vessel level measurement during Mode 5 reduced inventory p4
~
operations. The first system displays the output from the two inadequate core cooling probes which are located inside the reactor vessel. The range of these probes extends from the reactor vessel head to the 4jw9 fuel alignment plate (See Figure 2.8-3). The measurement of RCS water level via these probes is limited [h only to those periods when the reactor vessel head is installed.
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cagogo(mijYays the outpEt in the hot leg ,regett. This systemItsp trom two probes specifically designed with thermocouples clustered m the hot leg region (See Figure 2.8-4). The benetit of this design is that it permits very accurate measurement when the reactor vessel water level is in the hot legs. The HJTC systens compensate for the flow gradient across the core associated with the operation of only one SCS suction line. The HJTC instruments are located in areas which minimize the effect c,f the core outlet nozzles. The HJTC sensors have an accuracy and response time consistent with the maximum Approved Design Meterial. ProbabiKstic Risk Assessment Page 19.8A.47 C
System 80 + Design Control Docume D l CLUSTERED REACTOR THERMOCOUP VESSEL PROSE 3 FLANGE N UPPER HEAD l REGION S NN8HG k D ANDOWN l \ . l
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I 1 Schematic Representation of One Alternate Design for Narrow Range Figure 2.8-4 Heated Junction Thermocouple Probes Approved Design Matenal . Probabilistic olisk Assessment Page 19.8A 193
c FIGURE i ! g $"N I 1 a) Ta/ / ction, to nkConne/ / RC.S' Lc'op f! b) Tank Connection to RCS Loop , with Surge Line ' without Surge Line i
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i System 80+ DCD - Potential Design Channes Item Number: y7 Summary
Description:
Reduce Safety Depressurization System (SDS) Line Size ! Affected DCD Sections: CDM: None ADM: Figure 5.1.2-3 l Chapter 6.7 l Chapter 5.4 i Change
Description:
During the Design Certification engineering, a preliminary design analysis for the SDS line size was performed that showed 6 (six) inch piping and valves would pass more than the minimum
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flow to meet the EPRI requirements to mitigate a Total Loss of Feedwater (TLOFW) event. i During detailed engineering, a more detailed system calculation was performed to determine if 4 (four) inch piping and valves would pass adequate flow. The new analyses confirms that a 4 (four) inch system passes more than enough flow to meet the requirements specified in the DCD
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(Section 6.7) and preserves the validity of the original TLOFW analysis. This change will revise the RCS/SDS P&lD (DCD Figure 5.1.2-3) to revise the SDS piping and valves from 6 to 4
- mehes. Modifications to the line and valve sizes are also required in Section 5.4 and 6.7.
1 Using a four inch SDS system for future System 80+ contracts will allow the use of the same gate and globe valves used for Yonggwang 3 & 4. These valves have been tested and qualified for this service, so proven components would be used and retesting would not be required. DCD hbrkups Attached? Yes i l l 1 t TKSdedch. doc 8 6/11/96
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m _ _ _ . . . . _ _ _ _ ._. . _ _ _ _ -_ _ _ . _ , I l . System 80+ Des /an controlDocument Table 5,4.10-1 Pressurizer Parameters Propeny Parameter Design pressure, psia 2500 Design temperature, 'F 700 Normal operating pressure, psia 2250 Normal operating temperature 'F 652.7 I Intemal free volume, ft3 2400 Normal (full power) operating water volume, ft) .1200 l Normal (full power) steam volume, ft' 1234 Installed heater capacity, kW 2400 Hester type immersion Spray flow, mmimum design capacity, gpm 375 Bypass Spray flow, continuous, gpm 1-6 Noules Surge, in. (nominal) 12, schedule 160 l Spray, in. (nommal) 4, schedule 160 Safety valves, in. (nommal) 6, schedule 160 l Safety depressurization, in. (nominal) A// schedule 160 Instrument Level, in. (nommal) 3/4, schedule 160 Temperature, in. (nommal) 1 schedule 160 l Pressure, in. (nommal) 3/4, schedule 160 ! Hester, 0.D., in. 1-1/4 l l l l 4 i T } Approved Deeign Material. RCS and Connected Systems Page 6.4-46
System 80+ Design controlDocumen ! cylindrical in shape, about six inches in diameter and range in length from fourteen feet to sixteen feet from the junction with the horizontal header. Because of the asymmetry of the IRWST, the spargers on the side with the ICI chase are submerged about ten feet and about twelve feet on the opposite side. l The evaluation of the hydrodynamic loads was performed with methods and codes developed by ABB-Atom and verified through single cell and in-plant measurements. Analyses were completed based on the maximum mass rate of flow for four PSVs during rapid depressurization following the TLOFW event. l l l Figure 6.7-5 shows typical results (in this case for the header section) of the influence on the loads due to the gas dynamic shock, the water and air clearing phases and finally steam discharge. l The loads on the SDS piping are within the design capability of piping and supports for SRV piping. Thus, design of the SDS piping supports can utilize standard methods for piping analyses and support design. l l Table 6.7-1 Safety Depressurization System - Active Valve List l I l Power Source l Valve Number Type Line Size 125V DC Bus Actuator l l Reactor Coolant Gas Vent Valves RC-410 Globe 1.5 inch A Solenoid l RC-411 Globe 1.5 inch B Solenoid l RC-412 Globe 1.5 inch C Solenoid RC-413 Globe 1.5 inch D Solenoid RC-414 Globe 1 inch B Solenoid RC-415 Globe 1 inch A Solenoid l RC-416 Globe 1 inch D Solenoid RC-417 Globe 1 inch C Solenoid RC-418 Globe 1.5 inch A Solenoid l RC-419 Globe 1.5 inch B Solenoid 1 l Rapid Depressunzation Valves RC-408 Gate Yfinch B Motodil i RC-406 Globe l J/f/ inch D Motod!1 RC-409 Gate 1/ finch A Motod31 RC-407 Globe hfinch C Motod!1 til 480 VAC motor operator supplied from Class IE 125V DC through invener and step-up transformer. 1 Approved Design Material
- Engineered Safety Features Page 6. 7 17
f System 80+ DCD - Potential Design Changes Item 8: Addition of Alternative LBB Evaluation Method DCD Section: 3.6.3. 3.9A. Ficure 3.9A-12 Dfscrintion of Change: The NRC has approved a change for determining leak-before-break (LBB) crack stability criterion on load. The System 80+ certified crack stability criterion on load was that a pipe with , a leakage crack length subject to loads of 42 x (NOP + Maximum Design Load) was required to i have significant margin between the material and loading curves. An alternative stability criterion on load is 1 x (NOP + Maximum Design Load) when the components of the NOP load i as well as the NOP plus Maximum Design loads are combined by the absolute summation l method. This criterion was approved by the NRC staff on another ALWR design. The System i 80+ DCD will be changed to include this criterion as part of an acceptable alternative method for demonstrating LBB. l Related Channes to DCD l l l DCD Section Insert # Revision 3.6.3.7 3.6.3.7 1 An acceptable alternative method for the margin on loads and margin on crack length evaluations is to combine each component of the NOP l load and the Maximum Design Load absolutely. This method is referred to as 'the absolute summation of loads method'. If this l attemative method is used, the margin on load for the leakage crack size i is reduced from 42 to 1. The margin on crack length (2 time the l leakage crack size) remains the same. l 3.6.3.8, 3.6.3.8 Alternatively, cracks of the length that leak at the rate given above can (2nd bullet) withstand the absolute combination of normal operation load components and maximum design load with afactor of1. App 3.9A, 3.9A.1 1 If the absolute summation of loads method is used to evaluate the 11.9.6.5.1 margin on load and margin on crack length, the PED for the ai analysis is cons *ructed using the formulas Me = (NOP i + SSEi ) and Me = SF i for the points labeled "1" and Me = (NOP 2 + SSE2 ) and Me = SF 2 for the points labeled "2". App 3.9A, 3.9A.2 1 If the absolute summation of loads method is used, the PEDS will be
; 11.9.6.6 reconstmeted, and the piping design, evaluation and reconciliation will be based on the reconstructed PEDS.
Figure 3.9A-12 3.9A.3 W When the absolute summation ofloads method is used to evaluate the margin on load and the margin on crack length, ' Max. Load /V2' becomes ' Max. Load'. l
System 80+ Desian controlDocument l a leak detection capability of 1.0 gpm, with a safety margin of 10. The LBB eve.luation of the System 80+ main steam line inside containment is based on a leak detection capability of 1.0 gpm and a safety margin of 10. See Appendix 3.9A for further discussion of flow rate correlation. 3.6.3.4 Material Properties For the main coolant loop, the hot and cold leg piping material is SA516 Gr70 or SA508 CLI A. All hot-and cold-leg pipe-to-pipe welds and the pipe-to-reactor vessel, steam generator and reactor coolant pump welds are carbon steel. All main loop component nozzles are SA508 CL 1 A,2 or 3 or SAS41 CL 1,2 or 3. The surge line is SA312 Type 347 or Type 316 stainless steel, resulting in bimetallic safe end welds. He shutdown cooling line and the direct vessel safety injection line are Type 304.or 316 stainless steel. The main steam line is SA516 Gr70. He stainless steel piping fabricated for the surge, shutdown cooling and direct vessel injection lines are seamless pipes. The detailed analysis of cracks in pipe welds requires consideration of the properties of the pipe and the weld materials. Previous work by C-E has shown that a conservative bounding analysis results when the material stress-strain properties of the base metal (lower yield) and the fracture propenies of the weld (lower toughness) are used for the entire structure, (Reference 11). This material representation is applicable to all LBB analyses discussed in Section 1.9 of Appendix 3.9A. For both the final design and as-built configurations, material properties for piping systems subject to LBB which are listed in Section 3.6.3 will be reviewed. If either the base metal or the weld is found to have lower fracture toughness properties than those given in Appendix 3.9A, a LBB reanalysis using the material with the lower fracture toughness properties as the basis for the J-R curve will be performed. He tensile (stress-strain) curves and the Jo vs. Aa curves are required for each material type. Additional commitments with respect to review of final design and as-built configurations for piping systems subject to LBB are given in Section 1.9 of Appendix 3.9A. 3.6.3.5 I2akage Crack Length Determination It is necessary that hypothesized through-wall cracks open significantly to allow detection by normal leakage monitoring under normal full power loadings. The method for determining the appropriate leakage crack length is described in Section 1.9.6.2 of Appendix 3.9A. 3.6.3.6 Computation of J Integral Values 3.6.3.6.1 Range of Crack Sizes ne range of crack lengths are calculated using a detailed stability analysis of the through-wall cracks in the piping evaluated. The finite-element analysis is performed for the leakage crack size and twice that length. This procedure, therefore, considers the stability of a range of crack lengths for all locations selected for the analysis. 3.6.3.6.2 J-Integral j 1 ne stability of through-wall cracks is evaluated using the J-integral technique. The J-integral is i determined in the finite-element analysis for pressure, normal operation, and maximum design load, i Approved Design Meterial Design of SSC Pope 3.6 33 l il l
I I System 80 + Design ControlDocument which is the largest of the dynamic loads (due to safe shutdown earthquake, thermal stratified flow, rapid valve closure, or other load) included in the crack stability analysis. The J-integral is determined for two different crack lengths for each geometric model. For the margin on loads evaluation, the J-integral for the leakage crack size is evaluated for V2 x (Pressure + NOP - Maximum Design) loads. For the margin on crack length evaluation, the J-integral for 2 times the leakage crack size is evaluated for (Pressure + NOP + Maximum Design) loads. 3.6.3.7 Stability Evaluation The stability of the cracked pipes is assessed by comparing the J-integral value due to the applied loads on the pipe to the material crack resistance. The stability criterion for ductile crack extension employed is: if J-applied < J ci material, and (dJ/ja) , ppa,a < (dJ/da) .,,n i then crack stability is assured. The change in J-integral with crack length "a" is determined by analyzing several crack lengths in the region of interest. For a leakage crack of length "a", crack lengths "a", a-6, and a+6 are analyzed. Similarly, the change in J-integral with crack length in the region of length "2a" is determined by analyzing cracks with lengths 2a,2a-6, and 2at6. This method provides the derivative information in the two regions of interest. The variation of J with crack length in the region of "a" and "2a" is plotted along with the material curve. Evaluation of the plots allows for direct verification of the stability criteria. l The evaluations are performed for the locations chosen to envelop all limiting cases. The pipes with the leakage crack length subject to loads of V2 x (P + NOP + Maximum Design Load) and the pipes with crack length twice the leakage crack length with loads of (P -- NOP + Maximum Design Load) are g demonstrated to have significant margin between the material curve and the loading curve, indicating that py,rg,yall pipe locations satisfy the LBB crack stability criteria.
- ~ - - O- ~b"n See Appendix 3.9A, Sections 1.1.9.5.4 and 1.1.9.6 for a discussion of LBB design criteria development and a further discussion of analytical methods.
3.6.3.8 Results The piping listed in Section 3.6.3 and evaluated by the methods described above are shown to meet all the ((criteriafor application ofleak-before-break))2 according to NUREG 1061, Volume 3. Specifically,
~
these criteria ([ require thefollowing:))2
- Cracks which are assumed to grow through the pipe wall leak significantly while remaining stable. The amount of illeakage is detectable with a safety margin of at least afactor of10}}2
! unless otherwise justified. e (( Cracks of the length that leak at the rate given above can withstand normal operation plus maximum design load loads with a safetyfactor of at least 42.
/NSEA.7 -1. &
1 2 NRC Staff approval is required prior to implementing change in this information; see DCD Introduction Section 3.5. Asyweved oenign Material Design of SSC Page 3.6-34
1 l System 80 + Desian controiDocument ; I e Cracks twice as long as those addressed above will remain stable when subjected to normal operation plus maximum design load.}}2 Site specific evaluations will confirm that the bases for the LBB acceptance criteria are satisfied by the final as-built design and materials of the piping systems listed in Section 3.6.3 and will be documented l in a LBB evaluation report. References for Section 3.6
- 1. " Evaluation of Potential for Pipe Breaks," NUREG-1061, Vol. 3.
- 2. ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Class 1, 2 or 3.
- 3. ASME Code for Pressure Piping, B31, Power Piping, ANSI /ASME B31.1.
l 4 USNRC Branch Technical Position MEB 3-1 Rev. 2 - Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, attached to Standard Review Plan 3.6.2, June,1987. )
- 5. American National Standard Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture, ANSI /ANS 58.2-1988.
- 6. R. T. Lahey, Jr. and F. J. Moody, " Pipe Thrust and Jet Loads," The Thermal Hydraulics of a Boiling Water Nuclear Reactor, Section 9.2.3, pp. 375-409, Published by American Nuclear l
Society, Prepared by the Division of Technical Information, United States Energy Research and ! Development Administration,1977.
- 7. RELAP 4/ MOD 5, Computer Program User's Manual 098. 026-5.5.
l 8. USNRC Regulatory Guide 1.45 " Reactor Coolant Pressure Boundary Leakage Detection ! Systems." t
- 9. Not used
- 10. NUREG/CR-2781, " Evaluation of Water Hammer Events in Light Water Reactor Plants," July 1982.
l '
. 11. " Analysis of Cracked Pipe Weldments," EPRI NP-5057, February 1987.
- 12. USNRC Regulatory Guide 1.11 (Safety Guide 11), Instrument Lines Penetrating Primary Reactor Containment; including supplement, Backfitting Considerations.
- 13. Not used 2
NRC Staff approval is required prior to implementing change in this informationt see DCD Introduction Section 3.5. Anwoved Design Meterial . Design of SSC Page 3.6-35
[ l i 4 System 80 + Design Control Document l l In order to evaluate the derivative in the region of the leakage crack tip, three meshes are used. For a j given leakage crack length "1" and model crack length ai, the three meshes have crack lengths ai-d, ai, and ai rd. The value d is a length appropriate to the anticipated amount of stable crack growth. This is indicated in Figure 3.9A-26. These three meshes are used in the analysis of the leakage crack. l l Similarly, three more meshes are generated for the analysis of twice the leakage crack lengths 2ai-d,2ai, ' and 2ai+d. I For each load step in the analysis, the loading curve as a function of crack length is fit to a quadratic: 2 J(a) = Ci a + C 2a + C3 ! l l The values at a, a d provide the boundary conditions necessary to evaluate the constants Ci , C2, and C3. At each loading point, the function is differentiated. This provides the dJ/da values for the loading curve. The material curves J(a), dJ(a)/da are evaluated at increasing crack extension. The loading
- functions J(a), d1(a)/da are evaluated at either a, or 2ai , whichever crack length is being evaluated. Each point on the J vs. df/da loading curve corresponds to a different load state. As long as the loading curve stays below the material curve, Jwe < Jm7 and ;
i dJt .om/da < djur/da the crack growth is stable. For the case of increasing load, the loading curve will eventually iutersect l the material curve. At this point the crack will experience unstable crack growth. At this point of i instability, Jwe = Jm7 and dJww/da = dJmr/da Development of the J versus dJ/da diagram for determining points of instability is shown in Figure 3.9A-27. 1.9.6.5 LBB Piping Evaluation Plots 1.9,6.5.1 Constructing an LBB Piping Evaluation Diagram The method by which LBB Piping Evaluation Diagrams.(PEDS) are constructed allows for the evaluation i of the piping system in advance of the final piping anal sis, incorporating LBB considerations into the piping design. The LBB PED for each specific pipe size, material and pressure is prepared prior to the piping design and analysis and is used to evaluate critical points in the pipeline. The PED is constnicted to allow the maximum design load p'us, dead weight at the maximum design stress location to be plotted vs. the NOP load at the maximum design stress location. The maximum design load at any time during the plant operation is the loading used in the stability I analysis. Traditionally, this loading has been NOP+SSE. However, the combination of the NOP load ( and the largest of the design loads (i.e., the maximum design load) is used in the stability analysis (see l Section 1.4 of this appendix). In the case of the surge line, for example, the line is evaluated for the larger of either NOP+SSE and for Stratified Flow (SF). For the discussion that follows, the maximum design load is considered to be the SSE load, and the loading combination is NOP+SSE. Approved Design Material Design of SSC Page 3.%31
- .- ~ - - - - - _
System 80+ Design Control Document The LBB piping evaluation plot requires performing two complete LBB evaluations. The evaluations are for two NOP loads which span the typical loadings for the line under consideration. A completed typical diagram is shown in Figure 3.9A-28. The procedure used for generating that figure is as follows: )
- 1. Choose NOP = Pressure + NOP 3
- 2. Determine ai
- 3. Increase the analysis moment until the critical moment is found for ai and 2ai
- 4. Separate the critical analysis moment, M , into the correct addition of SSE and NOPi proportion for the ai and 2ai evaluations.
M= e g (NOP + SSE i) 3 (ai Analysis) M*' SSE =3
-NOP i and 6
I M = (NOP + SSEg) 3 (2ai Analysis) SSE = M, - NOP3
- 5. Plot SSE values at NOPi for the ai and 2ai analyses, respectively. This corresponds to the points labeled "1" in Figure 3.9A-28.
i
- 6. Repeat steps (1) to (5) for NOP2 . The results are shown in Figure 3.9A-28, labeled "2".
Two stability evaluations are performed for each pipeline under consideration in order to complete the piping evaluation diagram. When stratified flow (SF) is a critical thermal transient that must be considered in the stability analyses, the PED is constructed using the following relationships for the ai and 2a evaluations in (4) above:
- 4. M, = g SF i (at Analysis)
M SFi = - l and d M, = SFi (2ai Analysis) SFi = M, As in the case for SSE, Step (4) is repeated to determine SF 2 for the a sand 2ai analyses, and SF is O lotted vs. NOP.
/N%d i M, /} * /
JApneroved Design Meterial Design of SSC Pope 3.9A-32
I i l System 80+ Design control Document 1.9.6.5.2 Using an LBB Piping Evaluation Diagram Once the lines marking the acceptable areas of allowable piping loads are plotted as described in the previous section, normal operating piping loads and corresponding maximum design loads for the critical piping locations are plotted on the evaluation diagram. The critical locations are selected as the highest stressed point for each different type of material in the line. Figure 3.9A-29 shows how the plot is used for a hypothetical line. In this example, three points failed LBB and one point passed LBB. The reasons for each failure are given in the figure. The piping design can then be revised using the results (e.g., lowering the SSE response load by rerouting or by adding a snubber). Further review may result in other options for reducing the loads. I 1.9.6.6 Results Piping Evaluation Diagrams (PEDS) for piping systems listed in Section 3.6.3 are shown in Figures 3.9A-l 30 to 3.9A-36 and provide LBB acceptance criteria for these piping systems. These criteria are based on piping design parameters given in Table 3.9A-2. Analyses of preliminary design of these piping systems have demonstrated that the LBB criteria are met. Site-specific information will demonstrate that the final detailed design parameters of each piping system are consistent with those given in Table 3.9A-2 and that the final detailed design meets the LBB criteria of Figures 3.9A-30 to 3.9A-36. If design parameters for a piping system are not enveloped by those in l Table 3.9A-2, a new PED for that piping system will be constructed using the methodology given in this appendix and the piping design will be revised, as necessary, to meet the LBB criteria of the new PED. If a PED given in Figures 3.9A-30 to 3.9A-36 is applicable to the detailed design of a piping system but the detailed design does not meet the LBB criteria of the PED, the design will be revised to meet the LBB criteria of the PED. Reconciliation of the as-built piping systems with the final design will be documented in a LBB l Evaluation Report. The LBB Evaluation Report shall contain results of the LBB evaluations for as-built l piping. The LBB evaluations shall employ methods described in Section 1.9 of this appendix. Reconciliation of each as-built piping system qualified for LBB will be made by demonstrating that:
- 1. the as-built piping system meets the screening criteria of Section 3.6.3,
- 2. the dimensional and material properties of the as-built piping system are consistent with the parameters used in the development of the final LBB PED (s) for that piping system, and
'. 3. the as-built piping responses meet the ASME' Code allowables and the final LBB PED criteria.
See also Sections 3.6 and 3.6.3.8. L \ r
/NSB&T ~
- 3. 9. A . 2-Approvenf Design nietenel Design of SSC Page 3.BA 33
i l l
- e System 80+ Design Control Document I
i SELECT SELECT SELECT l PIPE LEAK DETECTION PIPE
)
! SIZE SYSTEM CAPABILITY 1 MATERIAL lf 10 m LDS CAPA8ILITY ! SELECT U SELECT LOW = e HIGH NOP LOADS NOP LOADS U y DETECTABLE DETECTABLE LEAKAGE LEAKAGE CRACK CRACK (DLC) U1f U U DLC 2 x DLC l UV U U LOAD LoAo J4 Ja CURVES CURVES VU VU MATERIAL J . dJ/tla -
~ J.a -
J dJ/da CURVES CURVES CURVES UV U1f O LBB PIPING MAX. LOAD /ff ) EVALUATION c MAXIMbM LOAD DIAGRAM NOP LOAD vs. MAXIMUM DESIGN LOAD 1 [
, /NSE27 '
i LBB Design Criteria Development Diagram Figure 3.9A-12 Approved Design Meterial. Design of SSC f*** #'##'# '
~ ~
System 80+ DCD - Potential Design Changes I Item Number: JPG Summarv
Description:
Eliminate Charging Pump Interlock (Add 160 gpm Flow Limit) Affcried DCD Sections: CDM: Section 2,7.16 ADM: Section 9.3.4 Channe
Description:
Reason for Change: The current System 80+ CVCS design includes an interlock in the charging pump controls so that both charging pumps cannot be operated at the same time. The interlock was added to limit the maximum charging flow within the upper bound assumed in boron dilution analyses. During System 80+ detailed engineering, a concern was identified related to charging pump transfer from the operating to the standby pump. Normally, plant operations personnel start a standby pump, then shut down the running pump. This method of pump switching ensures continuity of charging flow, especially seal injection flow to the reactor coolant pumps. It also prevents thermal transients on the charging and letdown subsystems. Although the dedicated seal injection pump (DSIP) could be started during charging pump switching, the DSIP does not have sufficient capacity to provide charging flow also. During the transition period without charging flow, letdown flow may be reduced or isolated because the charging stream is not available to cool the letdown stream in the regenerative heat exchanger. ,
Description:
The interlock that prohibits the operation of both charging pumps wotdd be deleted. In its place, a commitment will be added to Tier i that the CVCS will limit charging flow to the RCS to 160 gpm. This flow limit will ensure that the maximum flow assumption in the boron dilution analyses is preserved and that the design flow for CVCS charging line components, typically 200 gpm, is not exceeded. The flow limit will most likely be implemented by adding a flow indicator controller and an isolation valve, in the combined charging pump discharge, which will close when a high flow rate is measured. DCD Markups Attached? Yes dedchgjr. doc 4 6/11/96
- - - - - . ~ - - . . - - - - . - - - - -..- --- - - _ _ . -
Svstem 80+ Deslan ControlDocument
^ All three tanks are located within a common dike structure designed to contain the maximum
.
- l combined liquid inventory in the tanks.
e ne dike structure will be designed to comply with applicable state and local regulations. l 9.3.4.2 System Description 9.3.4.2.1 System De normal reactor coolant flow path through the CVCS is indicated by tse heavy lines on the flow diagrams (Figure 9.3.4-1, Sheets 1 through 4). Design parameters for the major components are shown in Table 9.3.4 4. Normal operating parameters for the CVCS are listed in Table 9.3.4-5. Process flow data is shown in Table 9.3.44. Letdown flow from the RCS passes through the tube side of the regenerative heat exchanger where an initial temperature reduction takes place via heat transfer to cooler charging fluid on the shell side of the heat exchanger. He regenerative heat exchanger is designed to cool letdown flow to less than 450'F for all normal operations and to heat the charging flow by a minimum of 100*F. A final terui .us reduction to the purification subsystem operating temperamre is made by the letdown heat exchanger. The letdown heat exchanger is sized to cool inlet water from the maximum regenerative heat exchanger l l outlet temperature to 120'F (or lower) for most operating conditions. Both the letdown and the I regenerative heat exchangers are designed for full RCS pressure and both are located inside containment
'p Letdown fluid pressure is reduced from RCS pressure to the operating pressure of the purificmion
! subsystem in two stages. The first pressure reduction occurs at the letdown orifices and the second occurs at the letdown control valves located downstream of the orifices. The letdown ori6ces are located inside containment. The letdown orifices are sized to pass the maximum letdown flow at full RCS pressure with one control valve full open. De orifice provides the pressure reduction necessary to minimize erosion of the letdown control valve seating surfaces during normal RCS operations. A bypass valve around the orifices is provided for low pressure operations. De process flow is then filtered via the purification filter purified via a purification ion exchanger, and sprayed into the VCT, An excess hydrogen invento is maintained in the RCS by keeying a hydrogen overpnessure on the VCT co
& 6t/CS /d,$s Meximem cbgc3 140u) fo %@.5 fo less +% x qu . f> %0f ne charging pumps normally take suction frotn the VCT and discharge to the RCS. During normal
/
L -,6.i b v e/tf r.: ^- 7 operations, one charging pump is running and the other is in standby. Au hu suum 0..a u - sion p-- 7 h +. - :--; ; ; " during all modes of plant operation. One letdown and one charging pump flow control valve are normally selected for use. Seal injection water is supplied
- to the Reactor Coolant Pumps (RCPs) by diverting a portion of the charging flow just downstream of the charging pumps. His seal flow is then heated in the seal injection heat exchanger to approximately 125'F before filtering. Once the flow has been filtered, the seal injection fluid is distributed to the four RCPs. De undiverted charging fluid is sent to the regenerative heat exchanger where it is hearmri before l
injection into the RCS. I A chemical addition tank and a chemical addition metering pump are used ta transfer chemical additives to the charging line downstream of the seal injection takeoff connection. Sufficient connections exist between the CVCS and the IRWST to allow for purification, inventory adjustments, and boron adjustments to the contents of that tank. O ( I AnweredDenten neeeadal Amagery syntene (2Ans) Pnee s.S2s
l
* \
l System 80+ l oesian controlDocument O Analysis of inadvertent deboration events initiated during operationa.: modes 2 through 6 (defined in the technical specifications) were performed. For Mode 1 operation, the reactivity addition due to a boron dilution event is less limiting than the CEA withdrawai events. These analyses show that Mode 5 (cold shutdown) in the drained down configuration results in the shortest available time for detection and termination of the event. Therefore, the initial conditions and analysis parameters are chosen for the cold shutdown operational nuxle to minimize the interval from initiation of dilution to the time at which criticality is reached. His results in the least amount of time between detection and criticality. The following are the analysis assumptions for Mode 5:
- 1. De Technical Specification lower limit on shutdown margin for cold shutdown is assumed,5.75% 4.
- 2. The most adverse initial core condition would be for an initial K, corresponding to 5.75% 4 suberitical and assuming suberiticality is maintained by boron concentration only.
- 3. The cold reactor coolant volume, including only the volurres for Mode 5 drained, is 2
3,%1 ft . A conservatively low reactor coolant mass was assumed by using the cold RCS internal volume. Assuming the coolant temperature of 210*F, the Technical Specification upper limit for cold shutdown, the resulting mass is 237,185 lbm. mmum O 4. Or + gin; pm E r:d :: 5: r ' ; :: i:: ==!- x ;;._, L'^ g z. h cin b b;h;d S i ."C*. !b :::r, fer ==. ;d=%e analysis usedh, charging pamp flow rategof 160 gpm which corresponds to 22.26 lbm/sec.
.to beACb
- 5. The critical boron concentration with all rods in except the largest worth rod stuck out and the inverse boron worth are 814 ppm and 66 ppm /%4, respectively, including uncertainties. The initial boron concentration for the cold shutdown mode is found by adding the product of the inverse boron worth and the minimum shutdown margin (i.e.
5.75 percent) to the critical boron concentration. The resulting minimum initial boron concentration in Mode 5 is 1193 ppm. Hus, the change of boron concentration from 5.75% 4 suberitical to critical is 379 ppm. The parameters discussed above are summnized in Table 15.4.6-1.
- Results Using the above conservative. parameters in Equation (3), the minimum possible time interval to dilute from 5.75% 4 subcritical to criticality is 67 minutes. Utilizing only the redundant, qualified neutron flux alarm, this time period will assure detection of a boron dilution event at least 30 minutes prior to criticality. Boron dilution will then be terminated before loss of shutdown margin by the operator actions discussed in Section 15.4.6.2.
15.4.6.4 Conclusions O ne inadvertent deboration event will result in acceptable consequences. Sufficient time is available for the operator to detect and to terminate an inadvertent deboration event ifit occurs. Fuel integrity is not challenged during this event. ww9aan w A.wwwww _ a ts.4 f f
So,0PW n y N S q h m - / lo c.l7t? y s System 80+ oesian controt occument n Table 15.4,6-1 Assumpdons for the Inadvertent Deboration Analysis Parameter Assumptions Cold RCS VolumelU (mid-loop operation), ft 3 3,961 RCS Mass (mid-loop operanon), Ibm 237,185 Volumetric Charging Rate, gpm 160 ME---- Mass Charging Rate, Ibm /sec 22.3 Dilution Time Cnament, t, sec4 10650 Initial Boron Concentration - Co, ppm 1193 Critical Boron Concentration - C., ppm 814 ! l O,
~
i l Q (U Includes the reactor vessel up to the mid-plane of the hot legs, half of a single hd leg, half of two cold discharge legs rad a shutdown cooling system. AnneMDea(ps hinsand AeaMant A& W 16.4 21
i . i , o i System 80+ Desian Control Document j A i Table 2.7.16-1 Chemical and Volume Contiel System (Continued) I I Design Consmitment Inspections. Tests, Aanlyses Acceptance Criteria I i j 5. Valves with response 5. Testing ofloss of motive 5. These valves change ! positions indicated on power to these valves will position to the position l Figure 2.7.16-1 change be performed. indicated on Figure i position to that indicated 2.7.161 on loss of motive i on the Figure upon loss of power. i motive power. 6.a) The letdownline is 6.a) Testing will be performed 6.a) The two CVCS letdown
- isolated by a safety using a signal simnhtine isolation valves inside injection actuation signal an SIAS. containment close upon l
. (SIAS). receipt of a signal i j simulating an SIAS. l l 6.b) The RCP seal controlled 6.b) Testing will be performed 6.b) The RCP sealcontrolled ! bleedoffline is isolated by using a signal simulating a bleedoffline isolation l a containment spray CSAS. valves close upon receipt l actuation signal (CSAS), of a signal simulating a
, GM.
k of M M Aa bG1h.** l
- 7. ^
.cL P E..4d r 7. To will be perf6rmed 7. N % ^- j ;, .
- se d eHieamere e ne+ee by r - -; d ._, --d" . ._ , J d
- j. .m . y -y ,, charging pumisfaserthe ~ ' - - -
E. . ! c . u ~. .c .u . .u . _ - - yfa cycs w;w, 7ai,_ cp , we, l l b ,a h.m)h
.M. "'
k N M h O Y Y M. I PM* l d
- 8. Motor operated valves 8. Testing will be performed 8. Each MOV having an (MOVs) having an active to open. or close, or open active safety function l safety function will open, and also close, MOVs opens, or closes, or opens j or will close, or will open having an active safety and also closes.
- and also close, under function under
! differential pressure or preoperationaldifferentis.1 fluid flow conditions and pressure or fluid flow ! under temperature conditions and under conditions, temperamre conditions. l J
- 9. Check valves shown on 9. Testing will be performed 9. Each check valve shown Figure 2.7.161 will open, to open, or close, or open on Figure 2.7.16-1, l
l or will close, or will open and also close, check opens, or closes, or opens
- and also close, under valves shown on Figure and also closes. l i system pressure, fluid flow 2.7.16-1 under system
- conditions, or temperature preoperationalpressure, i conditions. '
fluid flow conditions, or i temperature conditions. i 10. Flow limiting orifices are 10. Inspection of the as-built 10. Each letdown line flow l provided in the letdown letdown orifices will be limiting orifice has a i line, performed. cross-sectional area not 4 greater than 0.01556 l square feet. i d 4 J cwetEsd Desyv Adseemet pope 2.74e
System 80+ Desian controlDocument The letdown line is isolated by a safety injection actuation signal (SIAS). The RCP controlled bleedof' line is isolated by a containment spray actuation signal (CSAS). im,imriod h emsMA +he nn mnre than n= ^~-ine pur p 2
- :- r Tlw CVCS lis.,, fr %maximan citq:r Inspections, Tests, Analyses, and Acceptance Criteria}s .P/&
Table 2.7.16-1 specifies the inspections, tests, analyses, and associated acceptance criteria for the Chemical and Volume Control System. l l C i l I I I 1 I e 1
,D ca.nw w u.m,w n~ men
O Item W: Increased Damping for Response Spectrum Pining Analyses DCD Section: 3.7. Table 3.7-1. Fleure 3.7-32 Description of Chance: The NRC has approved a change in the maximum allowable value of applied damping for piping analyses in wluch the uniform envelope response spectrum analysis method is used. The maximum allowable damping in the System 80+ DCD for this type of analysis is currently based l on ASME Code Case N-411-1, which allows the damping to vary from 5% for modes of l vibration up to and including 10 Hz to 2% for modes above 20 Hz. The NRC approved change allows 5% damping for all modes of vibration when the uniform envelope response spectrum j analysis method is used on piping. This change was approved by the NRC staff on another ALWR design. Damping values from Table 3.7-1, based on piping diameters s12 inches and >l2 inches, continue to be used when piping is analyzed by time history or multiple support input methods. Related Chanres to DCD DCD Section Insert # Revision 3.7.1.3 Revise as marked Figure 3.7-32 Delete Table 3.7-1 Add entry for 5% damping, delete Note [2]; renumber Note [3] to [2]. Last 3 entries to Table 3.7.-l should read as follows: ([ Piping (diameter s 12 inches) 2.0 Piping (diameter > 12 inches) 3.0 Piping (umform envelope response spectrum 5.0}} ( ) method ofanalysis) 91 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5.
, System 80 + Desian controlDocurnent 3.7.1.2 Design Time History Since the System 80+ Standard Design is designed for generic site conditions, for the time history method of analysis, the generic free-field ground surface time histories are used as control motions in the analyses. In the soil-structure interaction analyses, for each generic site, the corresponding two horizontal and one vertical time histories at the free-field ground surface are used with the SSI model of that site.
For the fixed-base analyses, the rock outcrop time histories are directly used as the control time histories. The response spectra at 2,5 and 7% damping of control motion CMS 1, and 1,2,5 and 7% damping of control motions CMS 2 and CMS 3 and the corresponding spectral ordinates of the matching time histories are shown in Figures 3.7-1 to 3.7-12. The Power Spectral Densities of all time histories are included in Section 2.5. Each time history that is used in the SSI and rock analyses contains 20.48 seconds. For the SSI analyses, a time step of 0.005 seconds is used. For the Nuclear Island rock analyses. a time step of 0.0025 seconds is used. For Category I structures not on the Nuclear Island a time step of 0.005 seconds is used for both SSI and j rock analyses, e < . - ' 4v ' afhrn a Nv< refviremeMr : 3.7.1.3 Critical Damping Values uneVo'.em enVe /op .5f- fCMum 'T f/onr<_. ) yjping anal ysis . / Damping values used for various nuclear safety-relaEd structures systems and components are based upon Regulatory Guide 1.61 orDME Ca Cs ': " ' ' (Oc F@= ." :'"r. These values are expressed . in percent of critical damping and are given in Table 3.7-1. ".= 9e ogc= :px.a eld J l
- 'yi E d N pipig, A ping-dx .uc bcxd c . Code C= .'! ' %
3.7.1.4 Supporting Media for Seismic Category I Structures Category I structures are founded directly on rock or competent soil. For the Nuclear Island the , foundation embedment depth for System 80+ standard plant is approximately 51 feet (Reference 7). The rock properties and the layering characteristics, including shear wave velocity, shear modulus, and density, are given in Section 2.5. The System 80+ Nuclear Island is designed for the range of soil conditions discussed in Section 2.5 and shown in Appendix 3.7B. ] 3.7.1.4.1 Soil Structure Interaction (SSI) l
)
1 Two different types of analysis methodologies are used for the seismic analyses for the Nuclear Island. For the fixed-base cases, modal superposition time history analyses are performed i using the three control motions (CMS 1, CMS 2, and CMS 3) corresponding to rock site conditions. When a structure is supported on soil, the SSI is taken into account by coupling the structural model with the soil medium. To accomplish this, the methodology of the computer program SASSI(System for Analysis ) of Soil Structure Interaction, Reference 6) is used. Detailed methodology and results of the SSI analysis i for the Nuclear Island are presented in Appendix 3.7B. The methodology for the soil structure interaction for the non-Nuclear Island structures is presented in Appendix 3.7C. - l i Pope 3.M O~n:Mmeead. DQ of SSC
4 4 System 80+
; Desian control Document 6 =
e
/
/
s i V I / 4 = l l U ,s t l c_ f3 l8 3 = l 1 i I
, _ l ec l l ** l l 1 1 I I l 1 1 I
, = 1 I I I I I I I I I I l l l l 0
0 10 20 100 PREQUENCY, HZ GENERAL NOTE: DdMPlNG INDEPENDENT OF PIPE DLAMETER. Damping Value for Sdemic Analysis of Piping y1gure 3.7-32 Amved Danien m. Deafgn of SSC page 2.7 70
4 36A lk System 80+ __ Design controlDocument Table 3.6-3 Illgh-Energy Lines Within Containment (Cont'd.) Line Operating Operating item Functional Pressure Temperature No. System Description (>275 psig) ( > 200' F) Figure 38 Reactor Coolant SG No. 2 RCS Loop Yes Yes 5.1.2-1 2B Drain Line to RDT (High-Energy to Isolation Valve RC-332) 39 Reactor Coolant SG No. 2 RCS Hot Yes Yes ' . 1.2-1 leg Drain Line to - RDT (High-Energy to I Isolation Valve l RC-215) i c u r.-c. l (. 40.r) Reactor-Coolant Pressurizer Relief ~ Line #1 to Pres'surizer Yes
^'
Yes-
/'\
5.'1.2:3 - i j Safety. Val've'RC-200 1
# - 3 ' i 41/ Reactor Coolant Pressurizer Relibf ' Yes Yes 5.1.2-3 Ot / / / line #2 to Pressurizer '
Ph!S a* AsA*A
- /
/ Safety Valve RC-201 '
/ ! /
, e '
g /42 Reactor Coolant Pressurizer Relief Yes,e Yes 5.1.2-3 g pp, / ,f Line #3 to Pressurizer / /
/
/ / Safet'y Valve RC-202 / / / /
43 Reactor' Coolant 'P'ressurizer Relief / Yes / Yes 5.1.2-3 /
, Line #4 to Pressurizer ,
/ / [
Safety Valve RC-203 q 44 Reactor Coolant Pressurizer Spray Line Yes Yes 5.1.2-1 & from Cold Leg Loop 5.1.2-3 1A to IA Spray Control Valve RC-100E 45 Reactor Coolant Pressurizer Spray Line . Yes Yes 5.1.2-1 &
; from Cold Leg Loop 5.1.2-3 IB to IB Spray Control Valve RC-100F 46 Reactor Coolant Pressurizer Spray Line Yes Yes 5.1.2-3 from Loop 1A Spray Control Valve RC-100E to Pressunzer Spray e
Common Header Appmved Design Matwint - Denian or ssc :p% ', Pege 16 42
l [
* [TBM ll ,
1 System 80+ Design ControlDocument l Table 3.6-3 High-Energy Lines Within Containment (Cont'd.) l l Line Operating Operating item Functional Pressure Temperature l No. System Description (>275 psig) ( > 200'F) Figure l 57 Reactor Coolant Direct Vessel Injection Yes Yes 5.1.2-1 & Connection #4 to SIS 6.3.2-I C Interior Check Valve SI-217 _ 58 Safety Depressunzation f $"ia.d Off cf b~
> Pressurizer Safatu Yes Yes 5.1.2-3 System / f '/ . . " C-2^: Ea
. line to RC-409 1 Rapid Depress. Line),
+
l 59 Safety .., W b~ Yes Yes 5.1.2-3
; Depressurization Pressurizer Safety i
System VaiwRS203C-
! lane
- to RC-408 I 'fRapid Depress. Line),
_ ~ . -- - 60 CVCS Letdown Line from Yes Yes 9.3.4-1, Loop 2B to 5.1.2-1 Regenerative Hx 61 CVCS Letdown Line from Yes Yes 9.3.4-1 Regenerative Hx to Letdown Hx 62 CVCS Letdown Line from Yes No 9.3.4-1 Letdown Hx to Containment Penetration 63 CVCS Charging Line from Yes No 9.3.4-1 Coritamment Pen to Regenerative Hx 64 CVCS Charging Line from Yes Yes 9.3.4-1
.. Regenerative Hx to RCS Ioop 2A 65 CVCS Auxiliary Spray Line Ye6 Yes 9.3.4-1, j to Pressunzer Spray 5.1.2-3
! Common Header 66 CVCS SCS Hx Shutdown Yes Yes 9.3.4-1 Purification Line Cont Pen Check Valve CH-304 to Letdown Hx l Appmed Duign htenal Duign d SSC jyy Pope 3.w I
TC 0 l* System 80 + Design ControLD;; ;,eam l +s c Table 3.9-2 Loading Combinations,ASME Code Class 1,2, a&& Ccap na 3 Component ( l ~- Condition Design Loadingtil Combination Design PD + DW l Level A (Normal)(2) pg79w Level B (Upset)t2j POhDW level C (Emergency) POhDWf'MDE _Ievel D (Faulted) PO+DM+'SSENDF U y/ U
\
JN sfz% e.J ash k"* l [1] Legend: l PD = Design pressure l PO = Operating pressure DW = Dead weight SSE = Safe Shutdown Earthquake DE = Dynamic system loadings associated with the emergency condition DF = Dynamic system loadings associated with pipe breaks (nct elimmated by leak before break analysis) [2] As required by ASME Code Section III, other loads, such as thermal transient, and thermal gradient require consideration in addition to the primary stress producing loads listed. SSE is considered in equipment fatigue evaluations in accordance with Section 3.7.3.2. [3] For piping, see Tables 3.9-10 and 3.9-11. l Approved Design Material Design of SSC Page 3.9-73 (gj )
I
.L"TC M l3 3
System 80+ Design Contro/ Document l
)
Table 4.2-3 C-E Poolside Fuel Inspection Program Summary (Reference 70) (Cont'd.) i Shutdown l Reactor Date/ Cycle Inspection Program Scepe[1] ! Yankee Rowe 1987/18 VE,UT,SRE l Millstone-2 1977/1 VE 1982/4 VE St. Lucie-2 1987/3 VE,UT i 1989/4 VE,UT,SRE ANO-2 1981/1 VE, DM, SRE on C-E/EPRI Test Bundles 1982/2 VE, DM
/ -l 1983/3 VE, DM, SRE on C-E/EPRI and DOE Test Bundles i
,-l 1985/4 VE, DM, SRE on DOE Test Bundles j 1986/5 VE, DM, UT i
g -l/ _ _1988/6 eM yE,'nu SRE on DOE Test Bundles w , 'Th ru i j San Onofre-2 1984/1 VE, DM ) 1985/2 VE, DM i 1987/3 VE,UT,GS,SRE l 1989/4 VE, UT, SRE, DM i San Onofre-3 1985/1 VE,UT i 1988/3 VE,UT,SRE f 1 Palo Verde-1 1987/1 VE, DM i 1989/2 VE, DM ; Palo Verde-2 1988/1 VE, DM Waterford-3 1988/1-2 VE,UT,SRE 01 VE Visual F2=mination GS Gamma-Scannmg CS Crud Sampling S Sipping UT Ultrasonic Testing SRE Disassembly and Single Rod Exammations DM Dimensional Measurements ) 7 Ol%') Approved Design Meterial- Reector ") -u!95) Page 4.2-76
O 4 System 80 + ' Desiqn ControlDocument Table 5.4.7-2 Shutdown Cooling System Failure Modes and Effects Analysis Symptoms and Local Effects inherent Including Dependent Compensating Provision Remarks and No. Name Failure Mode Cause Failures Method of Detection Other Effects
- 1) Shutdown Cooling Fails Closed Corrosion, Effective loss of one Low flow indication Parallel redundant Valve is normally Pump Suction mechanical shutdown cooling train F-302, F-305; periodic shutdown cooling path locked open isolation Valve binding, cooling testing SI-106, $1-107 operator error Fails Open Same as la) No effect on SCS operation Periodic testing Nonc required
- 2) Shutdown Cooling Fails to start Mechanical Effective loss of one SCS Low Nw indication Parallel redundant '
Pump 1,2 failure, train F-302, F-305; periodic shutdowr. cooling path electrical testing __ failure
- 3) Shutdown Cooling tails Open Corrosion, Effective loss of one Low flow indication Parallel redundant Valves are locked
( i Pump Discharge , Closed mechanical shutdown cooling pump F-302 F-305; periodic shutdown cooling path open; min. flow line Isolation Valve T binding, testing will provide the min. SI-578, SI-579 operator error flow required to protect the pump Fails Open Same as 3a) No effect on SCS operation Periodic testing Nonc required
- 4) Shutdown Cooling less of Cooling lasufficient Diminished ability of liigh temperature Parallel redundant IIcat Exchanger 1,2 component subsystem to provide RCS indication from T-302, shutdown cooling path cooling water heat removal T-305 flow, excessive fouling
- 5) Shutdown Cooling Fails Closed Corrosion, Delays use of affected SCS Valve position Parallel redundant Same as 3a) lfcat Exchanger mechanical train indicator; periodic shutdown cooling path Bypass Valve binding, testing SI-312, SI-313 electrical failure Q
Fails Open Mechanical Effective loss of one Valve position Parallel redundant failure, shutdown cooling path indicator; perimlic shutdown cooling path s electrical testing % failure Approved Design Matedel- RCS and Connected Systems Page 5.4-40 G(9(d) 9 O - - - - - - - - - - - - - - - - - - - O_- -
e ws System 80 + Design control Document 1 The boron recovery portion of the CVCS accepts letdown flow diverted from the VCT as a result of feed and bleed operations for shutdowns, startups, and boron dilution over core life. The diverted letdown , flow, which has passed through a purification filter and ion exchanger, also passes through the pre-holdup l ion exchanger. The pre-holdup ion exchanger retains cesium, lithium, and other ionic radionuclides with ! high efficiency. The process flow then passes through the gas stripper, where hydrogen and fission gases are removed with high efficiency; thus (1) precluding the buildup of explosive gas mixtures in the holdup tank and (2) minimizing the release of radioactive fission product gases in aerated vents or liquid discharges. The degassed liquid is automatically pumped from the gas stripper to the holdup tank. Reactor coolant quality water from valve and equipment leakoffs, drains, and reliefs within the containment is collected in the Reactor Drain Tank (RDT) and scheduled for batch processing. Recoverable reactor coolant quality water outside the containment from various equipment and valve leakoffs, reliefs, and drains is collected in the Equipment Drain Tank (EDT) and scheduled for batch processing. Reactor coolant collected in either of these tanks is periodically discharged by the reactor drain pumps through the reactor drain filter and pre-holdup ion exchanger, and processed in the same manner as diverted VCT flow, as described above. This liquid is also pumped to the holdup tank. When a sufficient volume accumulates in the holdup tank, it is pumped by a holdup pump to the boric ! acid concentrator, where the bottoms are concentrated to within the range of 4000 to 4400 ppm boron. j The boric acid concentrator bottoms are continuously monitored for proper boron concentration, and a normally pumped directly to the BAST. In the event that abnormal quantities of radionuclides are { present, the bottoms are discharged to the LWMS. The boric acid concentrator distillate passes through ! a boric acid condensate ion exchanger, where boric acid carryover is removed. The distillate is collected l in the RMWT for reuse in the plant. If recycle is not desired, the distillate is diverted to the LWMS. When the SCS is operational, a flow path through the CVCS can be established for purification. This is accomplished by diverting a prtion of the flow from the shutdown cooling heat exchanger to the letdown line upstream of the letdown heat exchanger. The flow then passes through the purification filter, purification ion exchanger, and' letdown strainergead 4 returned-to-the-suction of 6e chn+de'?--t l _ cooling pumps. u
- A D The /C,7 ntc_
When continuous degasification of the RCS is desiret -
. rom the inlet of the VCT to the gas stripper, bypassing the pre-holdup io is processed in the gas stripper and is then returned to the VCT via the -rfm Mcs g h ed, VCT hydrogen overpressure can be used to replace the hydrogen Zha p p. ing process. The charging pumps take suction from the VCT, raid rett :S. ; A makeup subsystem of the CVCS provides for chang u; m outuu concemrauon. Boron is initially added to the CVCS usmg the boric acid batching tank (BABT). Reactor makeup water is added to the BABT via the makeup supply header, and the fluid is heated by immersion heaters. Boric acid powder is added to the heated fluid while a mixer agitates the fluid. A boric acid concentration of as high as 12 weight petcent can be prepared. Electric immersion heaters maintain the temperature of the solution in the boric acid batching tank high enough to preclude precipitation. The concentrated boric acid solution in the BABT is drawn into the boric acid batching eductor and diluted by fluid being circulated from the BAST via the boric acid makeup pumps. The reactor makeup water pumps can also be used by taking suction from the reactor makeup water tank and pumping the water through the eductor to the BAST.
G Approved Design Metenal- AuniEary Systems (C,/ cyo > Page 9.330
~ -. __-. . . _ - . _ _ _ - . . - . - - - . _ - . - _ - - - - , rem llo System 80+ Desian Control Document The resulting concentration of the refueling pool and the RCS is between the lower operating boron concentration limitation of the IRWST (4000 ppm) and the maximum operating boron concentration of the IRWST (4400 ppm). Thus, the contents of the refueling pool can be returned directly to the IRWST prior to plant startup without hindering plant operations.
During refueling shutdowns, the reactor makeup water supply piping is continuously monitored via flow switch F-250. An alarm is annunciated if flow is detected in order to prevent dilution of the refueling pool. 9.3.4.3 Design Evaluation 9.3.4.3.1 Availability and Reliability ) l l A high degree of functional reliability is assured by providing standby components and by assuring' fail- ) safe responses for the most probable modes of failure. Redundancy is provided as follows: Comnonent Redundancy Purification and Deborating lon Exchangers Bree identical components Charging Pumps One operating, one in standby Charging Pump Flow Control Valves One operating and one parallel, standby valve letdown Control Valves One operating and one parallel, standby valve
- j. Boric Acid Makeup Pumps Two identical pumps in parallel, one operates i
on demand, one in standby l l Gas Stripper Package ne gas stripper package includes redundant ] standby pumps i RLkts a4 i Seal Injection Filters Two identical pumps in parallel, one 3 i = da===4, one in standby . Purification Filters Two identical pendin parallel, one h ' ! engieumand, one in standby l- Reactor Makeup Water Pumps Two identical pumps in parallel, one operates on' demand, one in standby , Boric Acid Concentrator The concentrator package includes redundant I standby pumps b In addition to component redundancy, it is possible to operate the CVCS in a manaer such that some components are bypassed. It is possible to transfer boric acid to the charging pump suction header by bypassing the VCT. The letdown filter, and the purification and deborating ion exchangers can be , bypassed. Controlled bleedoff flow can be routed to the RDT rather than the VCT. l Independet and redundant gravity feed lines from the BAST to the charging pump suction are provided to assure makeup. De charging pumps also have an alternate source of borated water from the spent fuel pool, which is maintained above 4000 ppm boron. W onnen mennet- Aummy s vu ma (wn) rene s.w
a TTc M r[ System 80 + Design Control Document Table 9.3.4-4 Principal Component Data Summary (Cont'd.) Reactor Makeup Water Pumps Quantity 2 Type Centnfugal Design pressure 200 psig Design temperature 200'F Rated head 300 ft. Normal flow 180 gpm Normal operating temperature 40-120* F NPSH required 15 ft Matenal in contact with pumped fluid Austemtic stainless steel Fluid Demmeralized water Code None lioldup Pumps Quantity 2 I Type Centnfugal l Design pressure 100 psig l Design temperature 200*F l Rated head 145 ft Nonnal flow 50 gpm Normal operating temperature 40-120* F NPSH required I 10 ft Matenals in contact with pumped fluid Austenitic stainless steel Fluid 2.5 wt % boric acid, manmum Code None Reactor Drain Pumps Quantity 2 Type Centnfugal Design pressure 200 psig Design temperature 200*F Rated head 145 ft Normal flow 50 gpm Normal operating temperature 120'F NPSH required 10 ft Matenals in contact with pumped fluid Austenitic stainless steel Fluid 2.5 wt % boric acid, manmum Code for fluid end ASME III, Class 3 Volume Control Tank Quantity 1 Type Vertical, cylindncal Intemal volume 5,800 gallons (approx) Design pressure, miemal 75 psig Design pressure, external 15 psig Normal operstmg temperature 120*F A l Normal operating pressure Jatsts ? 20- 6o psa, < T-- J Approvect Design Material- AumEary Systems (i,/% ) Page 9.3-74
I f 4 12.HI6 System 80+ oesian controlDocument e Adequate clearances shall be provided for inservice inspection of the ASME Boiler and Pressure b Vessel Code Section III, Class 2 portions of the main steam system piping, in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code. e Loop seals are not utilized in safety valve inlets. 10.3.2.3.2 Valves l 10.3.2.3.2.1 Main Steam Isolation Valves (MSIVs) and MSIV Bypass Valves e The valves are designed so that no damage due to excessive closure force is incurred during closure under design conditions. e Backseating of valve stems is provided when the valve is in the full open position. e Unrecovered pressure loss from valve inlet to valve outlet at rated flow with the valve full open does not exceed 3 psid. e The Main Steam isolation Valve (MSIV) in each main steam line is remotely operated and is l capable of maintaining tight shutoff under the main steam line pressure, temperature and flow resulting from the transient conditions associated with a pipe break in either direction of the valves, pO e The MSIV leak rate through a closed MSIV flow does not exceed 0.001% of nominal fim at . 1200 psia in the forward direction and does not exceed 0.1% of nominal flow at 1200 psia in the reverse direction, o The full open to close stroke time of the MSIV's under fully developed steam line break flow is 5 seconds or less upon receipt of a Main Steam Isolation Signal (MSIS). The full open to close stroke time of the MSIV's Bypass Valves is seconds or less upon receipt of a MSIS. I e The MSIVs are supported such that the valve body and actuator will not be distorted to such a degree that the valve cannot close or be displaced as a result of pipe break thrust loadings, o The MSIVs and the MSIV bypass valves are designed, fabricated and installed such that the requirements for In-service Testing and Inspection of ASME Section XI, Subsection IWV can be met. e- He provisions of General Design Criteria 57 for containment isolation valves are met. e ne MSIV is a fail close valve; upon receipt of a Main Steam Isolation Signal the MSIV closes automatically. l e ne MSIV bypass valve is a fail-close valve; upon receipt of a Main Steam Isolation Signal the MSIV bypass valve closes automatically. l e ne MSIVs and their supports and the MSIV bypass valves and their supports are designed to withstand loads arising from the various operating and design bases events as specified in Section l 3.9.3. Anume o.on menormo - steem and rowere- '0 3-7
%l%
l System 80+ Desian ControlDocument } l e To permit testing for pH and the existence of foreign substances, sample connections are provided in the steam line piping between the steam generator nozzles and equalization header, e During initial startup and during periods of unit shutdown, the tripping mechanisms for the main j steam isolation valves are tested for proper operation in accordance with the technical : specifications. The valves are periodically in-service tested for leakage and freedom of movement during plant operation in accordance with ASME Code Section XI, Subsection IWV. e The main steam safety valves are tested during initial startup or during shutdown operation by checking the actual lift and closing pressures of the valves in comparison to the required design opening and closing pressures in accordance with .ASME Code, Section XI, Subsection IWV. e ASME Code Section XI, Subsection IWV requirements for in-service testing and inspection of nuclear safety-related valves apply to the atmospheric dump and atmospheric dump isolation valves. e A test will be conducted to verify MSIV response to a simulated Main Steam Isolation Signal (MSIS).
- 1. The objective of the test is to verify the function of the MSIVs and to confirm the 5 second closing time required by Section 10.3.2.3.2.1.
- 2. Since steam pressure is normally required to operate the MSIVs, a supply of steam at
.O conditions comparable to main steam is a prerequisite, in addition to the completion of.-
U construction activities on the MSIVs and required support systems. l 3. . The test method consists of the application of a simulated MSIS to the controls of the i MSIV under tett, the recording of temperature and pressure parameters upstream and downstream of the valve seat, and the timing of the closure process from the receipt of signal to the instance of valve closure as indicated by the valve stem travel indicator.
- 4. Acceptance criteria are that the MSIV operate to shut in 5 seconds or less, in accordance with Section 10.3.2.3. The test must be related by calculation and manufacturer's shop or type testing to the design basis conditions required by the safety function, e A test will be conducted to verify MSIV bypass valve response to a simulated Main Steam Isolation Signal (MSIS):
- 1. The objective f the test is to verify the function of the MSIV bypass valves and to confirm the,1 cond closing time required by Section 10.3.2.3.2.1. l 3 .
- 2. Construction activities on the MSIV bypass valves and their required support systems
- must be complete as a prerequisite.
- 3. The test method consists of the application of a simulated MSIS to the controls of the
- MSIV bypass valve under test, the recording of temperature and pressure parameters l upstream and downstream of the valve seat, and the timing of the closure process from
! receipt of signal to the instance of valve closure as indicated by the valve stem travel l indicator. i l l Anprovent Dennon ACetenal- Steam amt Power Convermen ((ofcyQ Page 10.3-13
l g { System 80+ oesign controlDocument c \ l 4. Acceptance criteria are that the MSIV bypass valves operate to shut in conds or less, in accordance with Section 10.3.2.3. The test must be related b'y calculation and manufacturer's shop or type testing to the design basis conditions required by the safety function. 10.3.5 Secondary Water Chemistry 10.3.5.1 Chemistry Control Basis l Steam generator secondary side water chemistry control is accomplished by:
- Close control of the feedwater to limit the amount of impurities which can be introduced into the steam generator.
- Continuous blowdown of the steam generator to reduce the concentrating effects of the steam generator.
- Chemical addition to establish and maintain an environment which minimizes system corrosion.
- Pre-operational cleaning of the feedwater system. l 1
- Minimizing feedwater oxygen content prior to entry into the steam generator. I Secondary water chemistry is based on the zero solids treatment method. This method employs the use of volatile additives to maintain system pH and to scavenge dissolved oxygen which may be present in the feedwater.
A neutralizing amine is added to establish and maintain alkaline conditions in the feedtrain. Neutralizing amines which can be used for pH control are ammonia and morpholine. Ammonia should be used in plants employing condensate polishing to avoid resin fouling. Although the amines are volatile and will not concentrate in the steam generator, they will reach an equilibrium level which will establish an l alkaline condition in the steam generator. l Hydrazine is added to scavenge dissolved oxygen which may be present in the feedwater. Hydrazine also tends to promote the formation of a protective oxide layer on metal surfaces by keeping these layers in a reduced chemical state.
; Both the pH agent and hydrazine can be injected continuously at the discharge headers of the condensate l pumps or condensate demineralizer, if installed. These chemicals are added as necessary for chemistry control, and can also be added to the upper steam generator feed line when necessary.
Operating chemistry limits for secondNy-side steam generator water, feedwater and condensate as
- extracted from Reference 2 EPRI Report NP-6239, are given in Tables 10.3.5-1,10.3.5-2 and 10.3.5-3.
l The limits stated are divided into three groups: normal, abnormal and immediate shutdown. The limits provide high quality chemistry control and yet permit operating flexibility. The normal chemistry conditions can be maintained by any plant operating with little or no condenser leakage. The abnormal steam generator limits are suggested to permit operations with minor system fault conditions until the l affected component can be isolated and/or repaired. The immediate shutdown limits represent chemistry l conditions at which continued operation could result in severe steam generator corrosion damage. l Approved Design Metenal. Steam and Power Conversion ((oI%) Page 10.3-14
e [ Md l9 System 80+ -Desian controlDocument h Table 19.7.5.1-1 Components in Seismic Fault Tree Models
- . m am w w-y eu.. w c _ - we_ N A T1m 9 A 7. M - 7 N NAME l DESC l PROB l HCLPE' )
Initiator and Special Events / ! EQ-MTC Adverse Moderator Temperature Coefficient (MTC) 1.00E-02 / i EQ-PSV Primary Safety Valve (PSV) Fails to Resent / 1.35 ! EQ-PSVO \ Primary Safety Valves (PSVs) Fail to Open / 1.35
- EQATWS Seismically induced ATWS s
/ 0.74
- EQLOSP N Seismically Induced Loss of Site Power / 0.12 l EQMLOCA s Seismically Induced Medium LOCA / 2.59 EQRVR Seismically Induced LOCA in Excess of ECCS O.73 i ' Capacity /
EQSLOCA Seismically Induced Small LOCA / 0.90 ] EQTRANS Seismically laduced Transient 0.05 l Emergency Feedwater System (EFWS) / ABDZEFWPBRKR Seismically ~ Induced Failure of EFW Motor Pump 0.95 l i Circuit Breaker /, ' AHFFASCSLOCA Operator Fails to Perform Aggressive Secondary 6.40E-02 Cooldown for Small LOCA / AKPZEFWSCABLE Seismically Induced Failure of EFWS Power Cable 1.80 APAJEFWP-102 Motor-Driven Emergency Pump EFWP-102 Fails to 3.00E-03 p Start \ / APAJEFWP-104 Pump EFWP-104 Fails to 3.00E-03 Motor-Driven Start Emergency / \ APAKEFWP-102 Motor-Driven Emergency PumpEFWP-102 Fails to 3.60E-03 4 Run / \ APAKEFWP-104 Motor-Driven Emergency Pump EFWP-104 Fails to 3.60E-03
- Ru , s
, APAVEFWP-102 Subtrain IB. Unavailable Due to Maintenance 2.00E-03 APAVEFWP-104 Subtrain 2B Unavailable Due to Maintenance. 2.00E-03 APTJEFWP 101 EFWfTurbine Driven Pump EFWP 101 Fails to Start 1.50E-02 APTJEFWP-103 EEW Turbine Driven Pump EFWP-103 Fails to Start 1.50E-02 APTKEFWP-101 f EFW Turbine Driven Pump EFWP-101 Fails to 'N 7.20E-03
~
/ Operate \
APTKEFWP-103 / EFW Turbine Driven Pump EFWP-103 Fails to y.20E-03
/ Operate N _
APTVEFWP-101 Subtrain IA Unavailable Due to Maintenance 5.00E-03 APTVEFWP-103' Subtrain 2A Unavailable Due to Maintenance 5.00E-113 . APTXDP101-103 Common Cause Demand Failure of EFW Turbine 1.19E-03\
/ Pumps s APTZEEWP Seismically Induced Failure of Turbine-Driven EFW \ 2.38
/ Pump \
i O ^orzerwsries AQWZEFWDGROOM s i -ic tir 1 a c 4 raii er erws riai S Seismically Induced Failure of Wall Separating o*o 0.84
/ EFWS/DG Room \
Anuwweef Deaign Metenal Prebebdseic Risk Assessment Pope 19.7-3
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> Mdh (9 System 80+ Design ControlDocument I
Table 19.7.5.4-7 Summary of HCLPFs for Seismic Sequences for Soil Site B4 Ol 4 T a Tcp aTi w* 6 / B L A aJ G ,,, q ,5
, g , 7, g, , ,; -,ge & gg, 7. :,4 7 1 Sequence HCLPF Dominant Cutset Mixed IICLPF Failure Prob. Dominant Cutset/
l EQSTR .73 - 1.25 2.50E-01 -
/
EQA-15 - -
.74 1.00E-02 1 SEIS-SBO .89 26 .12 2.10E-04 100 EQLP-8 .89 44 .89 5.00E-02 ,151 EQA-10 .89 6 .97 1.19E-03 / 42 EQLP-4 .90 61 - -
/ -
EQLP-7 .90 4 - - / - l EQSLO-3 .90 N 71 - -
/ -
EQSLO-6 .90 7 - - EQA-3 .90 80 - - EQA-4 .90 76. - - - EQSLO-9 .90 10 - - - - EQLP-3 .90 24 - l EQT-3 .90 80 - - - l EQT-7 .90 3 s
/ - - l 1
EQA-7 .90 2 \/ 3.25E-02 58 I EQA-8 .90 67 /\ 3.25E-02 174 EQSLO-10 .90 17 / \ 3.30E-03 64 EQT-4 .90 100 / \ 5.60E-02 138 EQT-8 .90 76 / \ 5.60E-02 142
.91 \
EQRVR -
/ - - -
EOLP-5 1.35 4 . EQSLO-4 1.35 6 - - - EQA-5 1.35 2r - - N - EQT-5 1.35 /5 - - N - EQA-12 1.35 / 2 - . N - EQA-14 1.35 / 1 - -
\ -
.- Ej',A-9 1.35 , 3 9.15E-03 \ 79 EOLP-9 1.35/ 5 9.15E-03 B2 EQ T-9 135 6 9.15E-03 71 \
EQiLO-ll / 1.35 11 6.40E-02 80 \ EQSLO-7 1.35 3 9.15E-03 17 \ l EQA-13 1.35 1 5.60E-02 27 \ EQLP-1.2 1.35 1 5.60E-02 53
- EQLP/II 1.35 1 1.55E-03 15 k EQML-3 2.59 1 5.60E-02 29 EQML-2 2.59 1 5.60E-02 17 Appmved Design Material ProbabiEstic Risk Assessment Page 19.7-26 l.
l 2 Tre+t z.O System 80+ Design ControlDocument i Figure 19.11.5.4.5.1-6 shows that the cavity basemat erosion is insignificant (less than 1 inch) during l l Q(h this transient. This is due to adequate quenching of the core debris in the reactor cavity. ! l l t l 19.11.5.4.5.3 Fission Product Releases l l j A summary of fission product group concentrations in the containment atmosphere at 24 hours after vessel breach is provided in Table 19.11.5.4.5.1-3.
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19.11.5.4.6 V Sequence l The dominant System 80+ V Sequence consists of an intersystem LOCA (ISL) initiated from a full shear break in the 16" diameter SCS line occurring within the containment building subsphere. This event is identified in the PRA as PDS 17. W , In this event all ECCS systems are operable. The failure of the SCS pipe outside of containment results in a gradual transfer of ECCS inventory from the containment to the subsphere. This ultimately results in failure of the ECCS function due to the unavailability of a water source. Details l of this transient are discussed below. l 19.11.5.4.6.1 RCS Response Characteristics l bey ser, 77CC l The ISL represents a large LOCA initiated d'utside of containment. Consequently the RCS response is l similar to that of the large LOCA discuss'ed in Section 19.11.5.4.2. In this case the SCS line break is equivalent to 1.4 square feet. The la In this I uf event core initiany-uncoveff ih[ger failure area results in a more rapid Rl seconds (See for= example Figure 19.11.5.4.6.1-2 and Table , I 19.1)o.5.4.6-1). whI'"d SIT @Mga-rapidly 4miM%cos da wrF et % ; (no F) The ECCS maintains the RCS covered until the IRWST is depleted and suction is lost to sECCS pumps. 'A =ond =nained ces usccvcry begins a 2 hwis." Support plate failure occurs at 13,800 l seconds and RV failure is predicted to occur shortly thereafter. The large failure area results in a rapid system depressurization to near atmospheric pressure which is sustained for the duration of the transient (Figure 19.11.5.4.6.1-1). H A summary of key transient parameters is provided in Table 19.11.5.4.6-2. b l 19.11.5.4.6.2 Containment Response Characteristics
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The ISLOCA releases all the .RCS and containment liquid inventory into the building subsphere. l Once the RV fails the corium is assumed to fully drop into the dry reactor cavity. Core concrete attack begins immediately. Concrete erosion will ultimately lead to a basemat failure. However, the bypass pathway provides a more direct means for releasing fission products to the environment. These MAAP analyses do not credit the water accumulation expected in the subsphere ECCS rooms to scrub fission products leaving the RCS. Furthermore detailed revolatilization models including the large length of SCS piping are likewise not considered in this demonstration. 19.11.5.4.6.3 Fission Product Releases {,h l V MAAP predicted fission product releases for the V sequence are summarized in Table 19.11.5.4.6-3. H l [,/96) Approved Design Matenal ProbabiUstic Risk Assessment Page 19.11-145 Q -W195t}}