ML20195C793
| ML20195C793 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 10/31/1988 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19297H195 | List: |
| References | |
| NUDOCS 8811030398 | |
| Download: ML20195C793 (14) | |
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. WESTINGHOUSE CLASS 3 E.
j; AMENONENT 2 TO RESAR-SP/90 PDA MODULE 4 REACTOR COOLANT SYSTEM 4
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O WEC P.O. Box 355 Pittsburgh, PA 15230 WAPWR-RCS AMENDMENT 2 5333e:Id OCTOBER, 1988 0011030398 881026 PDR ADOCK 05000601 A PDC ,
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AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 4 REACTOR COOLANT SYSTEM INSTRUCTION SHEET Replace current page 1.6-3 with revised page 1.6-3.
Replace current pages 1.8-22 through 1.8-25 with revised pages 1.8-22 through 1.8-25.
Replace current pages 3.2-1 through 3.2-3 with revised pages 3.2-1 through 3.2-3.
Replace current page 17.0-1 with revised page 17.0-1 Insert pe.ges A2-1 through A2-6 in Question / Answer section.
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O O WAPWR-RCS AMENDMENT 2 5333e:1d OCTOBER, 1988
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TABLE 1.6-1 (cont) ,
, NATERIAL INCORPORATED BY REFERENCE OC 3'Ei
- , Westinghouse SAR
- o. g Topical Revision Section Submitted Review Report No. Title Number Reference to the NRC Status WCAP-8301(P) LOCA-IV Program: Loss-of-Coolant Rev 0 15.0. 15.6 7/12/74 AE WCAP-8305 Transient Analysis WCAP-8302(P) SATAN-IV Program: Comprehensive Rev 0 15.0, 15.6 7/12/74 AE WCAP-8306 Space-Time Dependent Analysis of Loss-of-Coolant WCAP-8324-A Control of Delta Ferrite in Rev 0 5.2 6/23/75 A Austenitic Stainless Steel Weldments WCAP-8370 Westinghouse ESBU/NFBU Quality Rev 11 1.9, 178 10/06/88 U 2
.- Assurance Plan WCAP-8424 Evaluation of Loss-of-Flow Accidents Rev 1 15.3 5/30/75 U Caused by Power System Frequency Transients in Westinghouse PWRs WCAP-8510 Method for Fracture Mechanics Rev 0 5.3 7/76 U Analysis of Nuclear Reactor Vessels Under Severe Thermal Transients WCAP-8567-P(P) Improved Thermal Design Rev 0 15.0 7/75 A WCAP-8568 Procedure WCAP-8693 Delta Ferrite in Production Rev 0 5.2 3/16/76 B Austenitic Stair.less Steel Weldments R
g WCAP-8768 Safety-Related Research and Rev 2 5.4 10.78 B kg Development for Westinghouse Press-
.= g urized Water Reactors Program
_g Summaries - Winter 1977 g- through Summer 1978 m ro
i TABLE 1.8-2(continued)
O REGULATORY GUIDE 1.124, REVISION 1, JANUARY 1978,
- SERVICE LIMITS AND LOADING COMBINATIONS FOR CLASS 1 j t.!NEAR-TYPE COMPONENT SUPPORTS O The WAPWR SP/90 design will meet the intent of this regulatory guide.
i However, the SP/90 plant will be designed to conform to the rules of ASME III, Subsection NF, "Component Supports," 1986 Edition or to the latest code of ;
record.
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O WAPWR-ACS 1.8-22 AMENDMENT 2 OCTOBER, 1988 i
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i TABLE 1.8-2(continued)
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l O REGULATORY GUIDE 1.133 SEPTEMBER 1977, LOOSE-PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF LIGHT-WATER-COOLED REACTORS O Westinghouse has taken a position which takes exception to any need for regulatory guidance relative to loose parts monitoring. This position is O WAPWR-RCS 1.8-24 AMENDNENT 2 OCTOBER, 1988 3333e:1d
3.2 CLASSIFICATION OF STRUC(URES. COMPONENTS. AND SYSTEMS Certain structures, components, and systems of the RCS are important to safety because they:
- a. Assure the integrity of the reactor coolant pressure boundary.
- b. Assure the capability to shut down the reactor and maintain it in a safe condition.
O c. Assure the capability to prevent or mitigate the consequences of acci-dents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR 100.
- d. Contain or may contain radioactive material.
The purpose of this section is to classify structures, systems, and components according to the importance of the item in or, Jr to provide reasonable assur-ance that the facility can be operated without undue risk to the health and safety of the public. Table 3.2-1 of RESAR-SP/90 PDA Module 7' 2
"Structural / Equipment Design", delineates each of the items in the plant which l fall under the above-mentioned categories and the respective associated classification that the NRC, ANS, and industrial codes comittees have devel- ,
oped. Each of the classification categories in Table 3.2-1 is addressed in the following sectione. '
The classification of specific piping runs and valves in these runs is pro-vided in the RCS flow diagrams contained in this module. Instrumentation and !
electrical equipment required to shut down the plant er mitigate an accident which is associated with the RCS will be classified as 1E (or Safety Class 3 per ANS 51.1) and identified in the approp;iate module.
3.2.1 Seismic Classification Seismic classification criteria are set forth in 10 CFR 100 and supplemented by Regulatory Guide 1.29.
WAPWR-RCS 3.2-1 ANENDMENT 2 5333e:1d OCTOBER, 1988
TABLE 3.2-1 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR THE REACTOR COOLANT SYSTEM Principal Con-ANSI Safety Quality Code struction Codes Seismic 1 System / Component Location Class Assurance Class and Standards Category l
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Reactor. Vessel Steam Generators-Primary ,
Secondary Pressurizer (See Table 3.2-1 of RESAR-SP/90 PDA i Modulo 7. "Structural / Equipment Design") l Reactor Coolant Pumps l
Pressurizer Relief Tank Piping 2 Valves i
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O WAPWR-RCS 3.2-3 ANENDMENT 2 l
8333e:1d OCTOBER, 1988
, 17.0 QUALITY ASSURANCE 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION The Westinghouse Energy Systems Business Unit / Nuclear Fuel Business Unit Quality Assurance Program is described in Reference 1. 2 17.1.1 References f
- 1. ' Westinghouse Energy Systems Business Unit /Nu: lear Fuel Business Unit 2 ;
Quality Assurance Plan," WCAP-8370, Revision 11, October 1988. !
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WAPVR-RCS 17.0-1 AMENDMENT 2 3333e:1d DCTOBER, 1988 i I I
c RE0 VEST FOR ADDITIONAL INFORMATION WESTINGHOUSE ADVANCED PRESSURIZED WATER REACTOR (RESAR-SP/90)
DOCKET NO. 50-601 i
i The following Questions / Responses were formally transmitted in Addendum 2 to RESAR-SP/90 PDA in Westinghouse letter NS-NRC-88-3304, dated January 7, 1988. !
252.10 What is the basis (date, experiments, experience, etc.) for the use of Incoloy 800 as tube support plate material? (5.4.2.1, Module 4)
Response: ,
The steam generator tube support plate materials is Type 405 stainless !
steel and not Alloy 800. At the time of submittal of RESAR-SP/90 PDA Module 4, "Reactor Coolant System," a development program was underway which included examining the use of Alloy 800 for the steam generate- !
tube supports. The design was shown not to be feasible and the materials now will be Type 405 stainless. This change in design will be addressed in our FDA submittal.
. 252.11 What steps have been taken to avoid corrosion / erosion of J tubes attached to the feedwater ring? (5.4.2.1, Module 4)
Response
The feedring material i s carbon steel with a specified minimum chromium content of 0.08%. The J-tubes are Alloy 600 material. Thess i selections are to reduce the susceptibility to erosion / corrosion +
- compared to that shown for carbon steels with lower chromium content.
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The' following Questions / Responses were formally transmitted in Addendum 3 to 4
, O RESAR-SP/90 PDA in Westinghouss letter NS-NRC-88-3338, dated Way 13, 1988.
440.255 (Module 4, Section 5.2.2) Section 5.2.2 on page 5.2-3 states !
1 that the li relief valves of the residual heat removal i system (RHRS) arequid used to protect RCS at low temperatures when !
'O the RHRS is. in operation. Section 5.2.2.10 states that the pressurizer PORVs will be used for the low temperature i
i overpressure proter. tion (LTOP) function. Please clarify the i LTOP design for the WAPWR. l
- RESPONSE
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l l Low temperature overpressure protection for the SP/90 will be ;
- provided by the RHR suction relief valves. Please see response i to 440.256 .and modifications to Subsection 5.2.2.10 of j RESAR-SP/90 PDA Module 4, "Reactor Coolant System." l l^ l l 440.256 (Module 4, Section 5.2.2)' Expand this section to address the !
l assumptions used for a mass addition event relative to the LTOP !
l system design. j RESPONSE: !
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l The following has been included in modifications of Subsection l 5.2.2.10.2 of RESAR-SP/90 PDA Module 4, "Reactor Coolant ,
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j Systems." l l Westinghouse has performed evaluations to identify the relief j requirements for LTOP events and verify the acceptability of l using the RHR relief valves to protect the RCS. These j
! evaluations included: t
! I j o A preliminary determination of the RCS Appendix G limiting !
} pressure vs temperature O !
WAPWR-RCS A2-2 ANENDMENT 2 f 8333e:1d OCTOBER, 1988 !
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o Determination of the individual LTOP event mass / heat inputs and required relief valve relieving rates. Analyzed events include:
- a. a,e b.
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The bases for these analyses included:
- 1) Appendix G Limit - A low Appendix G pressure limit for the RCS was selected (( }) on which to base a e,c l
conservatively low set pressure for the kHR suction line reliof valves. The [ ] psig pressure limit is judged to a,c ,
be lower than the actual Appendix G limit that would be calculated with reactor vessel material containing a maximum
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, O 2) The set,,c,*r4 ior the RHR relief valves was established in accordance with Section 111 of the ASME Code, Part NC-7513.
The requirad capacity is based on the maximum RCS expansion rate determined as described in 3) and 4) below.
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! 3) The effects of flashing flow through the relief valve and/or 4
choked flow in the valve discharge line were evaluated assuming the RCS temperature was at 350'F, the maximum anticipated temperature when RHR provides LTOP protection.
l l 4) RCS expansion rates for the above two (a and b) mass input i events were determined at the nominal RHR relief valve set pressure. In the calculations the maximum allowed, as
, manufactured, pump head / flow delivery curves were used.
1O l WAPWR-RCS A2-3 AMENDMENT 2 8333e:1d OCTOBER, 1988 l
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- 5) RCS expansion rates for the above two (c and d) heat input events were conservatively determined as follows:
o The LOFTRAN ode was used to analyze the heat input and RCS expansion due to the inadvertent start of a reactor coolant puim during water solid operaU.on.
o The RCS expansior, due to inadve',%nt pressurizer heater operation considered the maxiK@ rate at which water could be displaced from the pressurizer by steam formation.
The results of this analysis show that using relief valves with the capacity of the current standard RHR suction relief valve, two of the four ISS RHR subsystems aligned to the RCS provide acceptable LTOP protection.
440.257 (Module 4, page 5.2-11) Item A states that to preclude O inadvertent ECCS actuation during heatup and cooldown, blockage of the safety injection signal actuation logic below 1975 psia is required. Discuss the impact of this design relative to a LOCA during modes 3 and 4.
RESPONSE
The initiation of a LOCA in modes 3 and 4 and the blockage of the safety injection signal actuation logic to prevent inadvertent ECCS initiation is currently being investigated generically for Westinghouse designed plants. Upon completion d of this generic investigation, the impact of the SP/90 design and the applicability of the generic conclusions relative to a LOCA in modes 3 or 4 will be applied to the SP/90 design in the FDA application. (Also see response to 440.256).
O O WAPWR RCS A2-4 AMENDMENT 2
. 5333e:1d CCTOBER, 1988
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The following Questions / Responses were formally transmitted in Addendum 5 to ;
. RESAR-SP/90 PDA in Westinghouse letter NS-NRC-88-3338, dated May 13, 1988.
t 440.21 Why does the credible mass input events only include the !
operation of two centrifugal charging pumps, with the normal [
letdown isolated? ,
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RESPONSE
4 Dur original response to 440.21 (Module 3) was unclear and
) inconsistent. To provide clarification to this response, our l "draft" response to 440.256 has been revised and Subsection l i 5.2.2.10 of RESAR-SP/90 PDA Nodule 4. "Reactor Coolant System" has been modified. The original response to 440.21 has been )
revised as tollows: "The responses to staff questions 440.255
- and 440.256 provide a discussion of the current SP/90 cold !
overpressure protection method, which utilizes two of four of i the ISS RHR suction relief valves during all low temperature !
operations."
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440.238 What is the desiga criteria used for sizing of the rupture disc i I
on the pressurizer relief tank? Is the rupture disc sized to [
accomodate all safety and PORVs lifting per the SRP? If not, e provide justification. -
RESPONSE
I Design criteria 0.) has been added to the desi.r bases contained .
i in Section 5.1.1 of RESAR-SP/90 PDA Nodule 4. "Reactor Coolant f 4 System." "The pressurizer relief tank rupture dites are [
4 designed to provide sufficient relief area to be consistent with [
j the combined relief capacity of both the pressurazer PORV's and l safety valves consistent with SRP requirements."
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WAPWR-RCS A2-5 AVENDMENT 2 ,
! B333e:1d OCTOBER, 1988
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j The following Questions / Responses were formally tr.nsmitted in .wtendun 6 to l
, RESAR-SP/90 PDA in Wertinghouse letter NS-NRC-88-3354, dated July 7, -1988. l
. t i 210.27 The staff's coments in 0210.25 and 210.35 als) apply to !
. portions of Table 1.8-2, Section 3.2.2, Section 3.2.3 and ,
l Section 5.2.2.6 of Nodule 4. These sections should be revised i i to agree with the response to Q210.35. l I I 2
RESPONSE: !
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Please refer to our original response to Staff Q210.1.
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- Westinghouse believes that the initiative taken to design the f i SP/90 plant to the latest industry codes and standards, includ- i l l l
ing ANSI /ANS 51.1, provides additional assurance that this plant ;
design will operate more safely and with better reliability than i current nuclear power plant designs. If this issue is not l settled prior to final design submittal, Westinghouse will ,
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! reexamine the manner in which safety- classifications are !
assigned for systems, components, and structures for the SP/90
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l The information in Table 1.8-2 which discusses exceptions taken
- 210.28 i to Regulatory Guides 1.124 Revision 1 and 1.130, Revision 1 ;
l does not completely con"orm the current staff positions relative ,
a to design criteria for ASt.E Class 1 cosponent supports. To be f i acceptable, this inforettion should be revised to provide a !
t commitment to construct all Class 1 component supports in !
I accordance with the rules of ASNE !!!, Subsection NF, "Component i
! Supports," 1986 Edition or to the Code of Record which will be !
applicable to the final WAPWR plant.
- (Reference Questions 210.60and210.65.)
j RESPONSE: j
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] Positions on Regulatory Guides 1.124 and 1.1; ' will be revised [
to state that the intent of the Regulatory Guides will be met.
- The final Westinghouse SP/90 design will cor?nre to the rules of ,
j ASME !!!, Subsection NF, "Component S t.& ts," 1966 Edition or (
i to the latest Code of Record, l
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WAPWR-RCS A2-6 AMENDNENT 2 i
L 3333e:1d DCTOBER, 1988
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