ML20237D572
| ML20237D572 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 12/17/1987 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Johnson W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 8712240040 | |
| Download: ML20237D572 (9) | |
Text
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December 17,.1987 g
'Do-ket No. 50-601 J
Mr..W. J. Johnson-Nuclear Safety Department, Westinghouse Electric _ Corporation Water' Reactor Division-Box 355 l
Pittsburgh,.PA 15230
Dear Mr.. Johnson:
SUBJECT:
- REQUEST FOR ADDITIONAL INFORMATION ON RESAR SP/90 As a result of our ongoing review of the RESAR SP/90 PDA application, we require _ additional information.in order to complete our review of the plant systems aspects of the design. Enclosed are review questions Q430.1-430.22.
Please respond to this: request within 90 days of the date of this letter.
If you have any questions regarding this matter, call ~me at (301) 492-8206.
Sincerely, original signed by Thomas J. Kenyon, Project Manager Standardization and.Nen-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special. Projects Office of Nuclear Reactor Regulation cc: See next page DISTRIBUTION:
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e Docket 1 No. : 50 601 3
Mr. W. : J. Johnson -
Nuclear Safety Department..
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-. Westinghouse Electric. Corporation Water Reactor. Division ^
~ Box 355 Pittsburgh, PA2 15230 E.
Dear Mr. Johnson:
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SUBJECT:
REQUEST FOR. ADDITIONAL'INFORMATION ON RESAR SP/90'
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As a result of our ongoing twlew of the RESAR SP/90 PDA application, we require additional information in order to complete our review of the plant systems aspects of the design. Enclosed are review questions Q430.1-450.22 j5
'Please respond to this request within 90 days of the date'of this letter.
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you have any. questions regarding this matter, call me at-(301) 492-8206.
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Sincerely,
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Thomas J. Kenyon, Project. Manager y~
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Reactor Project Directorate!
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Office of Nuclear Reactor Fbgulation
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December 17, 1987 Docket No. 50-601 Mr. W. J. Johnson Nuclear Safety Department Westinghouse Electric Corporation j
Water Reactor Division l
Box 355 Pittsburgh, PA 15230
Dear fir. Johnson:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON RESAR SP/90 As a result of i
,ngoing review of the RESAR SP/90 PDA application, we require additiont.1 information in order to complete our review of the plant systems aspects of the design.
Enclosed are review questions Q430.1-430.22.
Please respond to this request within 90 days of the date of this letter.
If you have any questions regarding this matter, call me at (301) 492-8206.
Sincerely, f
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T omas J.
enyon, Project Manager Standardization and Non-Fower
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Reactor Preject Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Rephtion ec: See next page I
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Docket No. STN 50-501
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RESAR-SP/90 q,
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Trevor Pratt Brookhaven National Laboratory Building 130 Upton, New York 11973 Mr. William Schivief, Westinghouse. Electric Corporation y
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i Mail Stop 4 0S Box 355.. <
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e, REQUEST FOR ADDITIONAL INFORMATION WESTINGHOUSE ADVANCED PRESSURIZED WATER REACTOR (RESAR SP-90) v DOCKET NO. 50-601 430.1 Which power supplies are used for the alarms on the watertight
~(3.4.1.2) access doors to rooms used to protect safety-related equipment from (Mod. 7) being adversely affected by flooding? Which power supplies are used on such alarms when they are on doors used to house components for more than one train of a single safety-related system? Confirm that redundancy in power supplies is provided as necessary in order to ensure that at least one train of safety-related equipment is protected from flooding.
1 430.2 Discuss any periodic tests or surveillance performed to assure that
-(3.4.1.2) the emergency floor drainage system is capable of preventing (Mod. 7) unacceptable water accumulation in safety-related equipment areas.
430.3 In subparagraph E of Section 3.5.1.1.2, you note the " remote" likeli-(3.5.1.1.2) hood of bonnets for valves rated at 600 psig and below from becoming (Mod. 7) missiles because of the low probability of simultaneous failure of the bonnet-to-body bolts. Provide information or data which shows i
that such failure is only a remote possibility.
Otherwise, bonnets (in valves rated at 4 600 psig) must be considered potential missiles and adequate protection must be provided against them.
430.4 In paragraph G of Section 3.5.1.1.2, you note that nuts, bolts, (3.5.1.1.2) nut and bolt combinations, and nut and stud combinations are not (Mod. 7) considered potential missiles because they have only a small amount of stored energy. Provide information which demonstrates that these potential missiles are not capable of damaging sensitive safety related equipment such as instrumentation or provide appropriate protection for such equipment.
430.5 In Section 3.5.1.1.2, you consider "certain vertical missiles
. (3.5.1.1.2) (although not considered credibic)" for which you state that (Mod. 7) pressurizer compartment coverall roof slab provides protection against potential damage to the containment, engineered safeguard components outside the pressurizer compartment.
It is difficult to determine how protection is provided for components near the pressurizer when these vertical missiles fall after striking the slab. Therefore, discuss how protection is provided for safety-i related components in the vicinity cf 2,he pressurizer from such an occurrence.
430.6 When discussing temperature and pressure sensors as a source of (3.5.1.2) potential missiles in Section 3.5.1.2, you conclude that the (Mod. 7) missile characteristics of these assemblies "are not of concern from a containment penetration standpoint." However, containment i
protection is not the only concern.
It must also be shown that potential missiles from these sources are not of concern to safety-related components or protection must be provided for them.
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. 430.7 In Section 3.5.4, " Missile Protection Interface Requirements," you (3.5.4) state that you have evaluated valves in high pressure systems outside (Mod. 7) containment within the NPB scope for potential missile sources. You have concluded that there are no credible missile sources from these valves "since there is no single failure associated with any potential valve parts that can result in the generation of a missile." Provide additional details concerning the design of valves such'as the presence of backseats, or special holddown devices which are capable of preventing missile generation. The discussion of missile generation should be revised to assume the potential generation of missiles from valves which do not have features to prevent missiles on failure of the component. The discussion should also address appropriate protection of safety related equipment from potential valve missiles.
We note that only in Section 3.5.1.2.3 is a potential method for propulsion of missiles mentioned, i.e., jet-propelled missiles. We
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find no mention of potential missile propulsion by means known as piston-type missiles sucn as those that result from failures in rotating equipment. Explain the omission of any discussion covering the means by which missiles are propelled especially since you provide formulas showing missile penetration of barriers (Section 3.5.3, " Barrier Design Procedures") but no discussion on the means by which the missile velocites used in those calculations may be determined. Confirm that all potential sources of missiles have been properly identified in accordance with the criteria of the SRP.
430.8 In Section 3.11.2.3, " Methods and Procedures for Environmental (3.11)
Justification," it is stated that qualification may be demonstrated by (Mod. 10) either type test, operating experience, analyses or a combination of these methods. However, 10 CFR 50.49 does not provide for qualification by analysis only. Therefore, it is the staff's position that qualification methods should be in.ompliance with paragraph (f) of 10 CFR 50.49. Your approach to equipment qualification should be revised accordingly.
430.9 With regard to the containment external pressure analysis, provide the (6.2.1) following information:
(Mod. 10) a)
The calculated and design containment external pressure.
b)
A description of the method and assumptions used to predict the containment external pressure resulting from inadvertent operation of the containment spray systems during normal plant operations.
c)
A description of the containment vacuum relief system (if provided).
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430.10 As described in SRP Section 6.2.1.1.A in order to satisfy the 1
(6.2.1) requirements of the GDC-38 to rapidly reduce the containment l
(Mod. 10) pressure, it is the staffs position that the containment pressure be reduced to less than 50% of the peak calculated pressure for i
the design basis loss-of-coolant accident within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the postulated accident. However, Figures 6.2-2, 6.2.3 and 6.2.5 show that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the postulated accidents, the containment i
pressure would be significantly higher than 50% of the calculated j
pressures. Provide justification for this deviation from staff guidelines.
j 430.11 Section 6.2.1.2.1.2 states that the effects of high energy line breaks 1
(6.2.1) in containment subcompartments will be analyzed in the Final Design (Mod.10) -Application (FDA) to establish criteria for the structural design of the compartment walls. Therefore, in the FDA, the following information should also be provided.
1.
For each compartment analyzed:
a)
Describe the moralization sensitivity studies performed to determine the minimum number of volume nodes required to conservatively preduct the maximum pressure for each sub-compartment. The nodalization sensitivity studies should l
include consideration of axially and radially within the l
subcompartment, particularly in the reactor cavity.
b)
Provide schematic drawings showing the nodalization of each j
subcompartment or compartment subdivision indicating nodal not free volumes and interconnecting flow path areas.
c)
Provide and justify the break type and area used in the
- analysis, d)
Provide and justify values of vent loss coefficients and/or i
friction factors used to calculate flow between model volumes e)
Discuss the manner in which movable obstructions to vent flow (such as insultation, ducting, plugs, and seals) were treated.
Include analytical justification if credit is 1
taken for the removal of such items to obtain vent area.
Provide assurance that vent areas will not be partially or completely plugged by displaced objects.
f)
Provide a table of blowdown mass flow rate and energy resulting in the highest differential pressure for each compartment.
g)
Provide the design differential pressure, peak calculated differential pressure, and time of peak pressure for each cc,mpartment.
l
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With regard to the method of analysis for subcompartments:
a)
Provide a description of the blowdown and pressure transient codes used in the analysis.
b)
Justify the blowdown model used showing that it maximizes the mass and energy release rates.
430.12 With regard to containment backpressure for ECCS analysis, Appendix K, (6.2.1)
"ECCS Evaluation Models," to 10 CFR Part 50 states, in part, that the (Mod. 10) containment pressure used for evaluating cooling effectiveness during reflood and spray cooling shall not exceed a pressure calculated conservatively for this purpose.
It further requires that the calculation include the effects of operation of all installed pressure reducing systems and processes. Branch Technical Position CSB 6-1, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation," provides additional guidance for the performance of a minimum containment pressure analysis and should be used when the analysis is performed. Therefore, justify the minimum containment pressure that has been used in the analysis of the ECCS to be conserva-tively low. Describe the conservatism in the assumptions of initial containment conditions, modeling of the containment heat sinks, heat sink surface area and any other parameter assumed in the analysis.
Provide the containment pressure temperature and sump temperature response for the most conservative assumptions.
430.13 With regard to the post-accident containment energy balance, Tables (6.2.1) 6.2-19 to 24 should be revised to show the energy removed / absorbed (Mod. 10) by each component (i.e., heat removal systems, passive heat sinks).
430.14 Provide an analysis (in accordance with Regulatory Guide 1.1) of the (6.2.1) available NPSH for the containment spray system pumps at the operating
. (Mod. 10) conditions during the post-accident injection and recirculation phases.
430.15 With regard to the containment fan cooler system, provide an analysis (6.2.2) to show the pressure differential across the ductwork and housing during (Mod. 10) accident conditions and verification that they have been designed to withstand the maximum pressure differential.
430.16 Section 6.2.5.2.2 of RESAR SP-90 indicates that the containment (6.2.5) hydrogen purge system would be addressed in Sections 6.2.2 and 9.4.6.
(Mod. 10) However, this system has not been discussed in the above sections.
Therefore, provide a discussion of the contaiment hydrogen purge system in accordance with the guidelines of Regulatory Guide 1.7.
430.17 Provide justification for the proposed air-lock testing schedule which (6.2.6) is a deviation from the requirements of Appendix J.
(Mod. 10) 430.18 Provide a discussion which indicates how TMI Action Plan Item II.F.1,
" Additional Accident Monitoring Instrumentation," requirements have been met.
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. 1 430.19 The proposed control room habitability system appears to satisfy (6.4) licensing criteria for a single unit. However, for a two unit site, (Mod. 13) the control room habitability system must meet the requirements of l
GDC 5 when the control room envelope is shared between both units.
Verify that a shared control room ventilation system for a two unit site will ensure control room insulation or emergency operation on demand assuming a single active failure when one unit is down for refueling / maintenance (such as diesel generator overhaul) while the other unit is operating. Otherwise, provide an interface criterion which specifies a separate control room and control room ventilation system for each unit.
430.20 In Section 10.4.7.5 " Instrument Applications," you state that "...the (10.4.7) feedwater controllers regulate the feedwater flow by continuously (Mod. 6&8) comparing the feedwater flow and steam generator water level with the programmed level and pressure-compensated steam flow signal."
Explain what happens when the following occurs:
a)
The level signal on one steam generator becomes defective and signals:
(1) A steam generator level higher than the control level with the actual level at the control level or below.
(2) A steam generator level lower than the control level with the actual level at the control level or higher.
i b)
The feedwater flow and compensated steam flow comparison instrumentation becomes defective and signals:
(1) An erroneous high steam flow rate while the feedwater I
flow rate is at the proper value or higher
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(2) An erroneous high stean flow rate while the feedwater flow 1
rate is at the proper value or higher 430.21 In the narrative for the EFW system, you note (Section 10.4.9.2.2.1, (10.4.9)
" Normal operation") that the startup feedwater system, SFWS, is (Mod. 6&8) automatically started following most transients. Further, you state in the next section (10.4.9.2.2.2, " Accident Operation"), that the l
i EFW system is automatically started in the event the SFWS fails to function properly during a transient or "...in the case of a more severe event." Describe the logic for the instrumentation and control system required to effect such action.
Indicate what is needed to discriminate between the normal and accident operation for the EFWS I
and SFWS to operate effectively. Verify that a single failure can not render the EFWS inoperable when required.
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430.22 EFW systems are required to comply with the criteria of Items II.E.1.1 (10.4.9) and II.E.1.2 of NUREG-0737. These criteria include fulfilling certain (Mod. 6&B) detailed recommendations, among them a reliability study which shows the EFW system unavailability to be within the range of 10-4 to 10-5/ demand. Provide the necessary information to permit the staff to confirm that the EFWS is in compliance with the criteria of NUREG-0737.
6 h.