ML20236V074
| ML20236V074 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 10/31/1987 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19302D109 | List: |
| References | |
| NUDOCS 8712040072 | |
| Download: ML20236V074 (13) | |
Text
{{#Wiki_filter:=r WESTINGHOUSE CLASS'3 4. AMENDMENT 1 TO RESAR-SP/90 PDA NODULE 13 - AUXILIARY SYSTEMS O O O
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O O : O P""** WAPWR-AS AMENDMENT 1 OCTOBER, 1987 5619e:1d i
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AMENDMENT 1 TO RESAR-SP/90 PDA MODULE 13 AUXILIARY SYSTEMS Instruction Sheet Replace current page 9.4-1 with revised page 9.4-1. Place remainder of package behind Questions / Answers Tab in Module 13. O l l l I I' O l O WAPWR-AS AMENDMENT 1 5619e:1d OCTOBER, 1987
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REQUEST FOR ADDITIONAL INFORMATION ON RESAR SP/90 Module 1 O Questions 281.1 and 281.2 are addressed in Amendment 2 to RESAR-SP/90 PDA Module 1, " Primary Side Safeguards Systems." l Module 13 281.3 Describe the test and inspection requirements for the chlorine (6.4) detectors, if needed, and the interface requirements, if any, of the chlorine detectors for the balance-of plant.
RESPONSE
The need for chlorine detectors for the Nuclear Power Block and balance-of plant, and number and location of any detectors is a design detail which will be established for the Final Design Application. Testing and inspection requirements will be l provided at that time, as applicable, consistent with current regulations. l 281.4 Describe the material compatibility and corrosion potential of (9.1.2) the 1% borated stainless steel, to be used for the construction , of the spent fuel rack modules, in the water chemistry and 4 radiation environment of the spent fuel pit.
RESPONSE
During the preliminary design phase of the RESAR-SP/90, Westinghouse was engaged in development work in the area of spent fuel racks including feasibility studies of new materials. Analyses and cost estimates performed for the 1% borated stainless steel, since the initial application of this O WAPWR-AS 281-1 AMENDMENT 1 5619e:1d OCTOBER, 1987
design in SP/90, have resulted in concerns with higher
-manufacturing costs and possible embrittlement in irradiated environments.
Until such time that the above concerns are addressed and shown ('] to be acceptable, Westinghouse is withdrawing the use of 1%
.V borated stainless steel and will provide a redesign of the spent fuel racks (Subsection 9.1.2) and new fuel storage area racks (Subsection 9.1.1) for the Final Design Application (FDA).
. f) V 281.5 Describe the sampling procedure, analytical instrumentation, and (9.1.3) frequency of analysis to monitor the spent fuel pit water purity and clarity, and the need for replacement of the demineralized resin and filters. List the chemical impurity and radiochemical limits.to be used in monitoring the spent fuel pit water and for initiating corrective actions.
RESPONSE
The sampling procedure, analytical instrumentation, and fre-(3 quency of analysis to monitor the spent fuel pit water purity and clarity, and resin and filter replacement, are detailed questions which will be addressed during the Final Design Approval phase. Chemical impurity limits for the spent fuel pit water will also be developed during the Final Design Approval phase; however, typical limits are Oiven in Attachment 281.5. 281.6 a. Describe the capability to take samples from the following O (9.3.2) locations during normal operation: chemical additive tank for the containment spray system, secondary-side water, steam generator blowdown, and gaseous radwaste storage tanks.
RESPONSE
The ability to sample the locations of interest is as follows: o Chemical Additive Tank: The chemical additive tank has been ! removed from the SP/90 containment spray system. WAPWR-AS 281-2 AMENDMENT 1 5619e:1d OCTOBER, 1987 _ - _ - _ _ - _ _ - - - - - - . 1
o Secondary Side Water: This is not within .the scope of the ,
- nuclear power block.
o Steam Generator Blowdown: This is not within the scope of the nuclear power block. i A* o ' Gaseous Radwaste Storage Tanks: The Gaseous Waste Processing- System uses charcoal absorption tanks, not compressed storage. Local sample connections are provided where appropriate throughout the waste systems. (]
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- b. Describe provisions to assure representative samples from the liquid and gaseous process streams, i.e., sampling line purging, plateout reduction in sample lines, and turbulent flow.
RESPONSE
To ensure representative samples, all sample lines will be p purged prior to collection of the sample. This purge flow will
\. be of sufficient velocity through the 3/8" sample lines to ensure turbulent flow and minimum crud accumulation or plate out in the sample lines. The purge time will be determined by the individual sample line arrangements. The purge flow can be routed to the CVCS volume control tank and reused, or .to the waste processing system for disposal.
- c. Describe the radiochemical analysis capability to quantify certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of reactor O
e core damage in a design basis accident. item II.B.3 in NUREG-0737). (Criterion 2 of
RESPONSE
I V The detailed radiochemical analysis capabilities of the plant will be developed during the Final Design Approval phase, and , will meet the requirements of NUREG-0737. O WAPWR-AS 281-3 AMENDMENT 1 5619e:1d OCTOBER, 1987
- d. Provide a procedure for estimating the degree of reactor n core damage based on measurement of fission product concen-Q trations, and taking into consideration of other parameters such as the readings from core exit thermocou-plant ;
i ples, water level indicators in the reactor- vessels, containment radiation monitors, and hydrogen analyzers. . O G
RESPONSE
The procedure for estimating the degree of reactor core damage post accident will be developed during the Final Design Approval phase, and will meet the requirements of NUREG-0737.
- e. Verify that reactor coolant and containment atmosphere sampling during post-accident conditions will not require an isolated auxiliary system (i.e., the let-down system) to be placed in operation in order to use the sampling system.
(Criterion 3 of item II.B.3 in NUREG-0737).
RESPONSE
p Reactor coolant and containment atmosphere sampling can be V accomplished post accident without placing any isolated auxiliary systems into service. Adequate sampling provisions are made for taking samples directly from the reactor coolant system and the containment atmosphere.
- f. Describe the capability of measuring dissolved gases in reactor coolant samples, and the capability of measuring dissolved oxygen if chloride concentration should exceed 0.15 ppm. (Criterion 4 of item II.B.3 in NUREG-0737).
RESPONSE
The detailed radiochemical analysis capabilities of the plant will be developed during the Final Design Approval phase, and will meet the requirements of NUREG-0737. Do WAPWR-AS 281-4 AMENDMENT 1 5619e:1d OCTOBER, 1987 ____._._m__ . _ _ _ _ _ _ . _
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- g. Describe the time needed for chloride analysis. (Criterion 5 of item II.B.3 in NUREG-0737).
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RESPONSE
The . detailed radiochemical analysis capabilities of the plant T will be developed during the Final Design Approval phase, and will meet the requirements of NUREG-0737. p h. Describe how. the radiation exposure guideline- of Section
-20.1(c) in 10 CFR Part 20 will be met in process sampling during normal operation.
RESPONSE
Radiation exposure to the operations during normal sampling will be maintained at very low levels by the use of remotely operated valves where appropriate, shielding, and adequate ventilation (including forced ventilation of the sampling hood to reduce the potential for airborne radioactivity exposure). A delay coil is O provided on the reactor coolant- system sample lines to allow I
' decay of the short lived radionuclides.
- i. Describe how the radiation exposure guideline of GDC-19 in Appendix A to 10 CFR Part 50 will be met during post-accident sampling. Consistent with the source term assumptions in Regulatory Guide 1.4, provide information on the predicted man-rem exposures based on person-motion for sampling, transport, and. analysis of all required parame-ters. (Criterion 6 of item II.B.3 in NUREG-0737).
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RESPONSE
Radiation exposure to the operators during post accident sampling will be kept within acceptable limits by the use of remotely operated valves, ventilation, shielding, and dilution of samples. Detailed post accident sampling procedures will be developed during the Final Design Approval phase. O WAPWR-AS 281-5 AMENDMENT 1 i 5619e:1d OCTOBER, 1987 m._ .__.._.______m._m_____._.______m ._m_ .___..._m__._.__.m _ ______ m________m_.-- __-
- j. Describe provisions to minimize personnel radiation exposure and counting errors in radiochemical analysis by dilution of O reactor coolant samples, radiation shielding, and ventila-tion to control airborne radioactivity. (Criterion 9 of item II.B.3 in NUREG-0737).
RESPONSE
O Radiation exposure to the operators during post accident sampling will be kept within acceptable limits by the use of remotely operated valves, ventilation, shielding, and dilution of samples. Detailed post accident sampling procedures will be developed during the Final Design Application phase.
- k. Describe provisions to filter the ventilation axhaust from the sampling station. (Criterion 11.b of item II.B.3 in NUREG-0737).
RESPONSE
The ventilation exhaust of the sampling station will be routed O to the plant HVAC system, where filtration will be provided. The filtration system will include roughing filters, HEPA filters, and charcoal adsorption units. Details of the ventilation system will be developed during the Final Design Application phase.
- 1. Describe the interface requirements of the process and post-accident sampling system for the balance-of plant.
RESPONSE
O The primary sampling system has no significant interface requirements with systems which are not part of the nuclear power block. O WAPWR-AS 281-6 AMENDMENT 1 5619e:1d OCTOBER, 1987 L_-_-_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ -
281.7. a. Describe provisions to monitor (a) the temperature upstream g (9.3.4, of the demineralizers to assure that the resin temperature 9.3.5) limits will not be exceeded, and (b) the pressure differen-tial across the domineralizers to assure that the pressure differential limits for the resin retention element will not be exceeded.
RESPONSE
For all domineralizers where any potential exists for exces-sively high temperature fluid, through failures' of other equip-ment, to be passed to the resin bed, temperature indication and automatic isolation valves are provided directly upstream of the resin beds. In particular: o The entire letdown line is isciated by closure of valve 8119-if high letdown temperature is detected at instrument T-130. This protects the CVCS mixed and cation bed demineralizers. o Valve TCV-1251 bypasses the recycle evaporator feed deminer-alizer on high temperature at T-1251 or high differential pressure at P-1252. All demineralized resin retention screens are designed to with-stand differential pressures in excess of normal operation or transient pressure drops through the beds. Local differential pressure indication is provided to allow monitoring of resin bed. condition.
- b. Describe provisions for automatically diverting or isolating O the flow stream to the demineralizers in the event the domineralizer influent temperature limits.
temperature exceeds the resin
RESPONSE
The provisions for automatically protecting the domineralizer resin from high temperature is described in the response to question part "a" above. WAPWR-AS 281-7 AMENDMENT 1 5619e:1d OCTOBER, 1987
- c. Provide a program to reduce leakage from the makeup and letdown lines in accordance with item III.D.1.1 of O NUREG-0737 and item III.D.1.1 of NUREG-0718.. q l
RESPONSE
1 Each plant specific applicant is responsible for providing a program to effectively reduce leakage from systems outside 1
- containment that would or could contain highly radioactive fluids during a serious transient. Submittal.would be required ,
for the Final Design Application (FDA) at least four (4) months prior to expected issuance of a full power license. Item III.D.1.1 of NUREG's - 0737 and - 0718 delineates the required steps the applicant should follow in identifying all systems outside containment which should be leak-tested to ensure ALARA ; levels of radiation. Note .that the RESAR-SP/90' letdown heat i exchangers are located inside containment and have no relief valves outside containment.
- d. Describe provisions to prevent precipitation of boric acid in the components and lines containing boric acid solutions, and .the adequacy of the system design to protect personnel from the effects of any toxic, irritating, or explosive chemicals that may be used.
RESPONSE
The APWR uses boric acid at a maximum concentration of four weight percent, with a precipitation temperature of approxi-mately 40*F. All safety related boration systems will use boric O acid at refueling concentration (less than two weight percent), which does not require any _ special concern for precipitation beyond freeze protection. All equipment and piping containing four weight percent boric acid will be located in heated rooms, or will have recirculation and/or heat tracing provided. l O WAPWR-AS 281-8 AMENDMENT 1 5619e:1d OCTOBER, 1987 _m._____m._m_ _ _ _ _ __-_
l 1 j-l ' Normal commercial standards for personnel protection from chemi-cal hazards will be followed; this will not be fully addressed O until the detailed design phase.
- e. Describe the interface requirements of the chemical and volume control system and the boron recycle system for the balance-of plant.
l RESPONSE: l The CVCS and BRS have no significant interface requirements with systems which are not part of the nuclear power block. NOTE: Due to a change to Table 1.9-3 of RESAR-SP/90 PDA Module 3,
" Introduction and Site," made in Amendment 2 (October, 1987) to ,
that module, a modification to page 9.4-1 of RESAR-SP/90 PDA l Module 13, " Auxiliary Systems" has been made to indicate a reference to design temperature parameters for the control room, and attendant areas and systems. O O WAPWR-AS 281-9 AMENDMENT 1 5619e:1d OCTOBER, 1987 i - - - - _ _ - _ _ - _ _ _ _ _ - l
l i s. F , ATTACHMENT 281.5 TYPICAL CHEMICAL IMPURITY LIMITS FOR SPENT FUEL PIT i O Item Specification l Solution pH Determined by concentration of. boric acid present. Expected range'is .. +(a,c) i Boric Acid as ppm B(a)
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Chloride +(a c) Fluoride , Makeup Water Shall meet reactor coolant makeup water specifications.(b) Calcium ") Magnesium
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- e. Determined by specific design of spent fuel -storage facilities. Expected range is from 2000 to 4000 ppm B. Value should be inserted in table for O reference.
Specification. This value can be obtained from the Plant Technical
- b. The dissolved oxygen limit may be waived.
O I O WAPWR-AS 281-10 AMENDMENT 1 5619e:1d OCTOBER, 1987
. 9.4 AIR CONDITIONING, HEATING, COOLING AND VENTILATION 9.4.1 Control EquipmentL Area Ventilation Systems The- control equipment area ventilation systems consist of the main control room air-handling system and the essential switchgear room air-handling 7 The_ control room system serves the habitability zone which contains systems.
the main control room and operator convenience' facilities. The switchgear room air-handling systems serve all the essential electrical equipment rooms including relay rooms, battery rooms, inverter rooms, and switchgear rooms. V . Refer to Subsection 6.4 of this module for specific control room information concerning habitability systems. . Refer to Table 1.9-3 of RESAR-SP/90 PDA-1 Module 3, " Introduction and Site" for design temperature parameters.
'9.4.1.1 ' Design Bases 9.4.1.1.1. Safety Design Bases i
SAFETY DESIGN BASIS ONE - The ' control room air-handlirig system is designed to maintain the control room air temperature below 85'F in all modes of plant operation for reliable operation of electrical and electronic equipment. SAFETY DESIGN BASIS TWO - The switchgear room air-handling system is designed to maintain temperatures in the essential electrical equipment rooms below 95'F during normal operating conditions. During post-accident conditions, the system will maintain temperatures in at least one set of rooms, either the A or'B rooms, below 104*F. SAFETY DESIGN BASIS THREE - The control room emergency circulation filter system is designed to provide protection against the effects of a postulated atmospheric release of radioactive materials or toxic gases. Dose to control room personnel shall be within the values of 10 CFR 50 Appendix A, GDC-19. O SAFETY DESIGN BASIS FOUR - The control equipment area ventilation systems l shall perform the required safety functions following a safe shutdown j earthquake, and shall withstand the effects of appropriate natural phenomena such as tornadoes, floods, and hurricanes (GDC-2). WAPWR-AS 9.4-1 AMENDMENT 1 5619e:1d OCTOBER, 1987
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