ML20209C598
| ML20209C598 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 11/08/1983 |
| From: | Johnston W Office of Nuclear Reactor Regulation, NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Novak T Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1473, TASK-2.B.3, TASK-TM NUDOCS 8311160284 | |
| Download: ML20209C598 (23) | |
Text
_
n
, , -) / ,
,c -
j' /I!L l
NOV 8 1983 Jocket No. 50-508 l MEMORANDUM FOR: Thomas M. Novak, Assistant Director for Licensing Division of Licensing I 1
FROM: William V. Johnston, Assistant Director l Materials, Chemical & Environmental Technology Division of Engineering
SUBJECT:
DRAFT SAFETY EVALUATION REPORT - WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT UNIT 3 (OL) DOCKET NO. 50-508 Plant Name: WNP-3 Suppliers: Combustion Engineering, Ebasco Licensing Stage: OL Docket No.: 50-508 Responsible Branch and Project Manager: LB #3; A. Veitti Reviewers: Frank Witt, B. Turovlin Description of Task: Operating License Review Status: Draft SER complete The Chemical Engineering Branch has reviewed Sections 5.4.2, 6.1.1, 6.1.2, 9.1.2, 9.1.3, 9.3.2, (including Item II.B.3 of NUREG-0737), 9.3.4, 10.3.5, 10.4.1, 10.4.6'and 10.4.8 of the FSAR through Amendment 3 against the criteria of NUREG-0800, (Standard Review Plan). Our draft Safe ~ty-Evaluation Report is enclosed.
- We have found Sections 5.4.2, 6.1.1, 6.1.2, 9.3.4, 10.4.1, 10.4.6 and 10.4.8 acceptable. We need additional information to complete our review of Sections 9.1.2 (Spent Fuel Storage), 9.1.3 (Spent Fuel Pool Cleanup System),
9.3.2 (Post-Accident Sampling), and 10.3.5 (Secondary Water Chemistry).
Our consultant, Brookhaven National Laboratory, under a technical assistance l contract, reviewed Sections 3.8.1, 3.8.2, 4.5.1, 4.5.2, 5.2.3, 5.4.2.6, i 6.1.1, 9.1.2, 9.2.1, 9.2.2, 10.3.6, 10.4.6 and 10.4.8 of this application.
! We agree with our consultant's conclusion that the materials compatibility l 1s adequate and acceptable.
By memorandum dated April 7, 1983, we have provided the Materials Engineering l Branch, secondary review input on FSAR Sections 4.5.1, 4.5.2, 5.2.3, 5.4.2.1, 6.1.1 and 10.3.6. We have concluded that the materials compatibility and corrosion aspects are acceptable.
Contacts: F. Witt B. Turovlin x28360 x28556 8311160284 831108 8 M ADOCK 050 Mee.
f';, j Thomas M. Novak ..
To meet the schedule for the SER the applicant should provide the requested information by March 1,1984.
William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering
Enclosure:
As stated cc: R. Vollmer D. Eisenhut V. Benaroya C. McCracken A. Veitti G. Knighton T. Sullivan S. Pawlicki J. Weeks (BNL)
F. Witt B. Turovlin DISTRIBUTION Central Files CMEB Reading CMEB Plant DE:CMEBll.l/,,dhCh D DE:CMEB DhA:t ET BTurovlin:bjp FWfti CMcCracken VBenaraya WVCohnston 11/3/83 11/,d/83 11/y/83 11/t-/83 11/$/83
Draft Safety Evaluation Report By the Office of Nuclear Reactor Regulation for Washington Public Power Supply System Nuclear Project No. 3 Docket No. 50-508 6.1.1 Post-Accident Emergency Cooling Water Chemistry INTRODUCTION This review is related to providing and maintaining the proper pH of the -
containment sump water and recirculated containment spray water following a design basis accident to reduce the likelihood of stress corrosion cracking of austenitic stainless steel.
The applicant will use borated water with a concentration of 4000-4400 ppm boron from the refueling water storage tank during the initial injection phase of containment spray.
The borated water will be mixed with a 40 percent by weight sodium hydroxide solution from the chemical storage tank.
The resulting solution will have a pH greater than 7, and will drain to the containment sump. Mixing is achieved as the solution is continuously recirculated from the sump to the containment spray nozzles during the recirculation phase of containment spray.
EVALUATION The post-accident emergency cooling water chemistry has been reviewed in accordance with Section 6.1.1 of Standard Review Plant (NUREG-0800, July 1981).
l We evaluated the pH of the water (mixture of refueling water storage tank and sodium hydroxide solution) in the containment sump. We verified by independent calculations that sufficient sodium hydroxide is available to raise the containment sump water pH above the minimum 7.0 level to reduce the probability of stress-corrosion cracking of austenitic stainless steel components. The removal effectiveness of the chemical additive for fission products in containment is reviewed in Section 6.5.2. We will review the surveillance requirements in the plant Technical Specifications to verify that sufficien?.
sodium hydroxide is maintained in the containment spray additive tank.
-mv
.' r, l
t CONCLUSION .
On the basis of the above evaluation, we conclude that the post-accident emergency cooling water chemistry meets the minimum pH acceptance criterion of Standard Review Plan Section 6.1.1, the positions of Branch Technical Position MTEB 6-1, the requirements of General Design Criterion 14 of Appendix A to 10 CFR 50, and the CESSAR interface requirements, and is therefore acceptable.
6.1.2 Organic Materials INTRODUCTION This evaluation is conducted to verify that protective coatings applied inside containment meet the testing requirements of ANSI N101.2,
" Protective Coatings (Paints) for Light Water Reactor Containment Facilities,"
American National Standards Institute (1972), and the quality assurance guidelines of RG 1.54 " Quality Assurance Req'uirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants." Compliance with these requirements provides assurance that the protective coatings will not fail under design-basis accident conditions and generate siijnificant quantities of solid debris that would adversely affect the engineered safety features.
EVALUATION We have reviewed the organic materials in accordance with SRP 6.1.2 (NUREG-0800). In the FSAR, the applicant states that the coating system
, used on exposed surfaces inside the containment have been qualified in accordance with ANSI N101.2. The applicant also states that the protective coating system for the containment are applied in accordance with RG 1.54.
The applicant meets the positions of RG 1.54 and the testing requirements of ANSI N101.2. These measures demonstrate their suitability to withstand a postulated design-basis a.ccident environment.
fm ,
'! )
The consequences of solid debris that can potentially be formed from unqualified paints are reviewed in Section 6.2.2. The control of combustible gases that can potentially be generated from the organic materials and from qualified and unqualified paints is reviewed in Section 6.2.5.
CONCLUSIONS On the basis of the above evaluation, we conclude that the o'rganic materials meet the testing requirements of ANSI N101.2 and the positions of RG 1.54 and are, therefore, acceptable.
9.1.2 Spent Fuel Storace INTRODUCTION ,'
Nuclear reactor plants include storage facilities for the wet storage of spent fuel assemblies. The safety function'of the spent fuel pool and storage racks is to maintain the spent fuel assemblies in a sub-critical array during all credible storage conditions. We have reviewed the compatibility and chemical stability of the materials (except the fuel l assemblies) wetted by the pool water.
EVALUATION The information provided in the FSAR was not sufficient for us to complete our evaluation. The applicant provided additional information by letters dated July 15 and September 2, 1983. The information provided in the applicant's responses is insufficient for the completion of our evaluation.
To complete our review, we need the following information;
. 1. Identify and list by either brand name, generic name (i.e. , S.S. Type
- 304,31'6) or industry specification all the materials used for fabrication of the high-density spent fuel storage racks and all other structural components wetted by cooling water, except the fuel assemblies, including the neutron poison material, rack leveling feet t
and rack frame. . - . - . - - _. -
~
(~ -
- 2. Provide test or operating data which provides assurance that the neutron poison material will not degrade during the lifetime of the spent fuel storage pool.
- 3. Provide a description of any materials monitoring program for the pool.
In particular provide information on the frequency of inspection and type of samples used in the monitoring program. ~
- 4. Show that no buildup of gases will occur in the cavities containing the poison materials.
9.1.3 Spent Fuel Pool Cleanup System INTRODUCTION .'
The spent fuel pool cleanup system is designed to maintain optical clarity and to remove corrosion products, fission products, and impurities from the spent fuel pool water. Water purity and clarity in the spent fuel pool, refueling pool, and refueling canal are maintained by filtering and demineralizing the pool water through a mixed bed
.demineralizer. The pool cleanup system consists of two parallel trains of cleanup equipment. The cleanup loop is normally run on an intermittent basis which is determined by the chemistry conditions of the spent fuel pool water.
. EVALUATION The information provided by the applicant was not sufficient for us to complete our evaluation. The applicant has not provided the additional information on the spent fuel pool cleanup system. l l
l
.~, .s
-S- ~
To complete our review, we need the following information:
Describe the samples and instrumentati.on and their frequency of measurement that will be performed to monitor the Spent Fuel Pool water purity and need for ion exchanger resin and filter replacement. State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water k
and for initiating corrective action. Provide the basis for establishing these limits. Your response should consider variables such as: gross gamma and iodine activity, demineralizer and/or filter differential pressure, demineralizer decontamination factor, pH and crud level.
9.3.2 Process and Post-Accident Sampling Systems A. Process Sampling. System INTRODUCTION The process sampling system is designed to provide representative liquid and gaseous samples drawn from the primary and secondary coolant systems, the associated auxiliary system process streams, and the spent fuel pool cleanup system. Provisions are made to assure that representative samples are obtained from well mixed streams or volumes of effluent by the selection of proper sampling procedures. In the event of an accident, all sample lines l which pass through the containment are automatically isolated by two fail-closed solenoid operated valves.
EVALUATION The information provided by the applicant has been reviewed in accordance with Section 9.3.2 of the Standard Review Plan (NUREG-0800, July 1981).
+ 1
(.
i
/ ,
l l
. The process sampling system includes piping and other components associated 4
with the system from the point of sample withdrawal from a fluid syst'em up to the analyzing sta. tion, sampling station, or local sampling point. Our review included the provisions proposed to sample all principal fluid process streams associated with plant operation and the applicant's proposed design of these systems, including the location of sampling points, as shown on piping and instrumentation diagrams.
We determined that the proposed process sampling system meets (1) the requirements of General Design Criterion 13 in Appendix A to 10 CFR Part 50 to monitor variables that can affect the fission process for normal operation, anticipated operational occurrences, and accident conditions, by sampling the reactor coolant, the ECCS core flooding tank, the refueling water storage tank, the boric acid mix tank, and the boron injection tank for boron
, concentration; (2) the requirements of General Design Criterion 14 in Appendix A to 10 CFR Part 50, to assure a low probability of abnormal leakage, rapidly propagating failure, and gross rupture, by sampling the reactor coolant
! and the secondary coolant for chemical impurities that can affect the reactor coolant pressure boundary material integrity; (3) the requirements of General Design Criterion 26 in Appendix A to 10 CFR Part 50 to control the rate of reactivity changes, by sampling the rcactor coolant, the refueling water
( storage tank, and the boric acid mix tank for baron concentration; and (4) the requirements of General Design Criteria 63 and 64 in Appendix A to l
10 CFR Part 50 to monitor for radioactivity that may be released from i
normal operations, including anticipated operational occurrences, and from postulated accidents, by sampling the reactor coolant, the pressurizer tank, the steam genarator blow-down, the secondary coolant condensate
- treatment waste, the sump inside containment, the containment atmosphere, the spent fuel pool, the gaseous radwaste storage tank for radioactivity, and the CESSAR interface requirements discussed in the CESSAR SER.
~
,m -
We further determined that the proposed process sampling system meets (a) the standards of ANSI N13.1-1969 for obtaining airborne radioactive samples; (b) the requirements of 10 CFR Part 20, 20.l(c) and regulatory positions 2.d(2), 2.f(3), 2.f(8), and 2.i(6) of Regulatory Guide 8.8, Revision 3, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably
, Achievable," to maintain radiation exposures to as low as is reasonably achievable, by providing (1) ventilation systems and gaseous radwaste treatment system to contain airborne radioactive materials; (2) liquid radwaste treatment system to contain radioactive material in fluids; (3)
' spent fuel pool cleanup system to remove radioactive contaminants in the spent fuel pool water; and (4) remotely operated containment isolation valves to limit reactor coolant loss in the event of rupture of a sampling line; (c) the requirements of General Design Criterion 50 in Appendix A to 10 CFR Part 50 to control the release of radioactive materials to the environment by providing isolation valves that will fail in the closed position; and (d) regulatory positions C.1, C.2, and C.3 of Regulatory Guide 1.26, Revision 3, " Quality Group Classifications and Standards for Water-Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants," and C.1, C.2, C.3, and C.4 of Regulatory Guide 1.29, Revision 3,
" Seismic Design Classification," by designing the sampling lines and components of the process sampling system to conform to the classification of the system to which each sampling line and component is connected, and thus meets the quality standards requirements of General Design Criterion 1 and the seismic requirements of General Design Criterion 2.
CONCLUSION On the basis of the above evaluation, we conclude that the proposed process sampling system meets the relevant requirements of 10 CFR Part 20, $20.l(c),
General Design Criteria 1, 2, 13, 14, 26, 60, 63, and 64 in Appendix A to 10 CFR Part 50, and the appropriate sections in Regulatory Guides 8.8, 1.26 and 1.29 and, therefore, is acceptable.
I
(~, .
, B. Post-Accident Sampling System (NUREG-0737, II.B.3)
~~
INTRODUCTION Subsequent to the TMI-2 incident, the need was recognized for an improved post-accident sampling system (PASS) to determine the extent of core degradation following a severe reactor accident. Criteria for an acceptable sampling and analysis system are specified in NUREG-0737, Item II.B.3.
The system should have the capability to obtain and quantitatively analyze reactor coolant and containment atmosphere samples without radiation exposure to any individual exceeding 5 rem to the whole body or 75 rem to the extremities (GDC-19) during and following an accident in which there is core degradation.
Materials to be analyzed and quantified include certain radionuclides that are indicators of severity of core damage (e.g., noble gases, isotopes of iodine and cesium, and' nonvolatile isotopes), hydrogen in the containment atmosphere and total dissolved gases or hydrogen, boron, and chloride in reactor coolant samples.
To comply with NUREG-0737, Item II.B.3, the applicant should (1) review and modify his sampling, chemical analysis, and radionuclide determination capabilities as necessary and (2) provide the staff with information pertaining to system design, analytical capabilities and procedures in l sufficient detail to demonstrate that the criteria are met.
EVALUATION l
l l By letter dated April 1, 1983, the applicant provided information on the PASS.
l Criterion (1):
The applicant shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The combined time allotted for sampling and analysis should be three hours or less from the time a decision is made to take a sample.
The applicant has designated that the static uninterruptable power supply be available for the post accident sampling system in the event of a loss of offsite power. Information was not provided on the capability to obtain and analyze reactor coolant and containment atmosphere samples within three hours from the time a decision is made to take a sample.
We find the applicant partially meets Criterion (1).
Criterion (2):
The applicant shall establish an onsite radiological and chemical analysis capability to provide, within the three- hour time frame established above, quantification of the following:
a) Certain radionuclides in the reactor coolant and containment atmosphere.t' hat may be indicators of the degree of core damage (e.g., noble gases, iodines and cesiums and non-volatile isotopes);
b) hydrogen levels in the containment atmosphere; c) dissolved gases (e.g., H 2
), chloride (time allotted for analysis subject to discussion below), and boron concen-tration of liquids; d) alternatively, have in-line monitoring capabilities to perform all or part of the above analyses.
The PASS provides diluted and undiluted liquid and gaseous samples for grab sample analysis, and in-line monitoring of hydrogen in the con-tainment atmosphere.
We find that these provisions partially meet Criterion (2). The appli-cant should provide a procedure to estimate the extent of core damage
m
)
based on radionuclide concentrations and taking into consideration other physical parameters such as core temperature data and sample location.
Also, information on an casite radiological and chemical analysis capa-bility must be provided.
Criterion (3):
Reactor coolant and containment atmosphere sampling during post-icident conditions shall not require an isolated auxiliary system
,e.g., the letdown system, reactor water cleanup system) to be placed in operation in order to use the sampling system.
Reactor coolant and containment atmosphere sampling during post-accident conditions does not require an isolated auxiliary system to be placed in operation in order to' perform the sampling function. The applicant's proposal to meet Criterion (3) is acceptable since PASS sampling is' performed without requiring operation of an isolated auxiliary system and all PASS valves which are not accessible after an accident are environmentally qualified.
Criterion (4):
Pressurized reactor coolant samples are not required if the applicant can quantify the amount of dissolved gases with unpressurized reactor coolant samples. The measurement of either total dissolved gases or H2 gas in reactor coolant samples is considered adequate. Measuring the 0 2concentration is recommended, but is not mandatory.
Pressurized reactor coolant samples are cooled and degassed to obtain representative dissolved hydrogen and oxygen samples at the PASS sampling station. We have determined that these provisions meet Criterion (4) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable.
r
~
- r. _. }
+
i Criterion (5): !
The time for a chloride analysis to be performed is dependent upon two factors: (a) if the plant's coolant water is seawater or brackish water, and (b) if there is only a single barrier between l primary containment systems and the cooling water. Under both of ,
the above conditions the applicant shall provide for a chloride -
analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample bein'g taken. For all other '
cases, the applicant shall provide for the analysis to be completed [
within 4 days. The chloride analysis does not have to be done !
onsite.
The applicant proposes that chloride analysis will be done at an offsite laboratory to meet the four day requirement. However, to comply with Criterion (5), specific arrangements need to be made with the offsite laboratory and a licensed shipping container needs tc be available for transporting the sample.
\
Criterion (6):
The design basis for plant equipment for reactor coolant and contain-ment atmosphere sampling and analysis must assume thac it is possible
, to obtain and analyze a sample without radiation exposures to any individual exceeding the criteria of GDC-19 (Appendix A, 10 CFR Part 50) (i.e., 5 rem whole body, 75 rem extremities). (Note that the design and operational review criterion was changed from the operational limits of 10 CFR Part 20 (NUREG-0578) to the G0C-19 criterion (October 30, 1979 letter from H. R. Denton to all licensees.))
The applicant has performed a shielding analysis on operator exposure while obtaining and tr'ansporting a PASS sample. This evaluation does
q ,
not meet Criterion (6) which requires a man-rem exposure estimate based on person motion study for sampling, transport, and analysis of all required parameters.
Criterion (7): l l
The analysis of primary coolant samples for boron is required for
'PWRs. (Note that Rev. 2 of Regulatory Guide 1.97 specifies the j
need for primary coolant boron analysis capability at BWR plants.)
The applicant will have the capability to analyze reactor coolant samples for boron. Boron accuracy is discussed in Criterion (10). This provi-sion meets the recommendations of Regulatory Guide 1.97, Rev. 2 and Criterion (7) and is, therefore, acceptable.
Criterion (8):
If in-line monitoring is used for any s'ampling and analytical capability specified herein, the applicant shall provide backup :
sampling through grab samples, and shall demonstrate the capa-bility of analyzitig the samples. Established planning for analysis at offsite facilities is acceptable. Equipment pro-vided for backup sampling shall be capable of providing at least one sample per day for 7 days following onset of the accident and at least one sample per week until the accident condition no longer exists.
Reactor coolant-dissolved oxygen and containment hydrogen concentration measurements will be performed in-line. The PASS has the capaibility of providing backup samples of diluted gas, diluted liquid and undiluted liquid.
We find these provisions meet Criterion (8) and are, therefore, acceptable.
Criterion (9):
The applicant',s radiological and chemical sample analysis capability shall include provisions to:
/ -[3-a) Ident'Ify.ind quantify the isotopes of the nuclide categories discus' sed above to levels corresponding to the source term given in Regulatory Guides 1.3 or 1.4 and 1.7. Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be provided. Sensitivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately 1p Ci/g to 10 Ci/g.
b) Restrict background levels of radiation in the radiological and chemical analyis facility from sources such that the. sample analysis will provide results with an acceptably small error (approximately a factor of 2). This can be accomplished through the use of sufficient shielding around samples and outside sources, and by the.dse of a ventilation system design which will control the presence of airborne radioactivity.
The radionucl. ides in both the primary coolant and the containment atmosphere will be identified and quantified. Provisions are available for diluted reactor goolant samples to minimize personnel exposure. The PASS can parform radioisotope analyses at the levels. corresponding to the source term given in Regulatory Guides 1.4, Rev. 2, and 1.7. Radiation background levels will be restricted by shielding. We find these provisions partially meet Criterion (9). The applicant should determine whether radiochemical analysis results can be obtained within an acceptably small error (approximately a factor of 2). Also, information on the ventilation system design provisions to control airborne radioactivty should be provided.
Criterion (10):
Accuracy, ranage, and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant sy' stems.
The applicant has not provided sufficient information for our review to determine comoliance with the requirements of Criterion (10). <
.. ; \
l
-14
, 4 l
Criterion (11): '
In the design of the post-accident sampling and analysis capability, consideration should be given to the following items:
a) Provisions for purging sample lines, for reducing plateout in sample line, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions to ifmit reactor coolant loss from a rupture of the sample line. The post-accident reactor coolant and containment atmosphere samples should be representative of the reactor coolant in the core area and the containment atmosphere following a transient or accident. The sample lines should be as short as pos'sible to minimize the volume of fluid to be taken from containment. The residues of sample collection should be returned
~
to containment or to a closed system.
b) The ventilation exhaust from the sampling station should be filtered with charcoal adsorbers and high-efficiency particulate air (HEPA) filters.
l The applicant has addressed provisions for purging to ensure samples are
- representative, size of sample line to limit reactor coolant loss from a rupture of the sample line, and ventilation exhaust from PASS filtered through charcoal adsorbers and HEPA filters. To limit iodine plateout, the containment air sample line is heat traced. We determined that these provisions meet Criterion (11) of Item II.B.3 of NUREG-0737, and are, therefore, acceptable.
i
~
k- . :, , ..
CONCLUSION -
On the basis of the above evaluatkon, we conclude that the post-accident sampling system meets five of the eleven criteria of NUREG-0737, Item II.B.3.
Additional information is needed to complete our review of the six remaining criteria:
Criterion (1):
Provide information on the capability to promptly obtain and analyze reactor coolant and containment atmosphere samples within three hours from the time a decision is made to take a sample.
Criterion (2):
Provide information on onsite radiological and chemical analysis capability. Provide a procedure to estimate the extent of core damage.
Criterion (5):
Make specific arrangements for chloride analysis at an offsite laboratory and also for a licensed shipping container.
Criterion (6):
Perform a person-motion study for man-rem exposures accumulated during sampling, transport, and analysis of all required parameters.
Criterion (9):
Discuss whether radiochemical analysis results can be obtained within
. an acceptably small error (approximately a factor of 2). Discuss the ventilation system design relative to airborne radioactivity control.
- m -
Criterion (10): .
Discuss the accuracy, range, and sensitivity of the PASS parameters to ensure an adequate description of the radiological and chemical status of the reactor coolant system.
9.3.4 Chemical and Volume Control' System Our evalua' tion of the Chemical and Volume Control System (CVCS) proposed for use in WNP-3 is presented in Section 9.3.4 of the CESSAR SLR. In that evaluation, we conclude that the CESSAR CVCS is acceptable, provided the CVCS interface requirements for balance of plant (EOP) are adequate.
CES!AR identifies interface requirements for the CVCS with the BOP in Section 9.3.4, which include.n'ormal and egergency power; protection from natur:41 phenomenhu s' ch as floods, winds, tornadoes; and earthquakes; protection from pipe failure and misslies; separation of components; thermal limitations; inspection and testing; materials cesp3tibiTity; system /cosp,7ent arrangements; radwaste management; overpressure protection; refueling water tani design parameters; alternate source of borated water from the spent (nel poul; fire protection; operating temperatura range:; environmental control; and mechanical interaction between componer.ts.
. I 10.3.5 Secondary Water Chemistry INTRODUCTION In late 1975, we incorporated provisions into the Standard Technical i Specifications that required limiting conditions for operation and surveillance requirements for secondary water chemistry parameters.
The Technical Specifications for all pressurized water reactor plants that have been issued an operating license from 1974 until 1979 contain either j . these provisions or a requirements to establish these provisions after j r l
l L
~
4 f ,/ ' )
_17 baseline chemistry conditions have been dctermined. The intent of the provisions was to provide added assurance that the operators of newly licensed plants would properly monitor and control seccadary tvater chemistry to limit corrosion of steam generator ccaponents such as tubes arid tube support plates.
In a number of instances, the Technical Specifications have significantly restricted tne operati'o nal flexibility of some p' ants with little or no benefit with regard to limiting tje;rsootion of steam generater tube ano the tube sup; ort plates. Basad on this experience and the kncwledge ga(ned it.
recent years, we have concluded th&t Techf.ical Specification limits are nut the most effective Way of essurif.g that steam generator degradatl6n tvill be minimized.
Due to the complexity *of the corrosion phenom 69a irvolved and the stata-of-the-&.'t at it ext:ts today, we are of the opinion th:t, in lieu of specifying l<mitirig canditions in the Technical Specification, a nioro effective apprcach 'would be to specify a Tedhnicai Speci'ication that required the innplem2ntation of a secondary water chemistry munitoring and control program containing appropriata proced.res ano s.dmini:trative controls. This has been the approach for contrcl of secondary ; vater programs since 1979.
The required program and procedues are to be developed by applicants with input from their reactor vendor or other consultants, to account for site and plant specific factors that affect water chemistry conditions in the stea,s generators. In our view, plant operation following sucn procedures would provide assucance that licensees would dovete proper attention to contro} ling secondary water chemistry, while also providing the needed flexibility to allov them to deal effectively with an off-normal condition that might arise.
EVALUATION In the FSAR the applicant provided details of a secondary water chemistry monitoring and control program. The information provided in the FSAR was
./'S ~
~
not sufficient for us to complete our evaluation. The applicant provided additional information by letters dated July 15 and September 2,1983. The information provided by the applicant is inadequate to complete our evaluation.
To complete our evaluation, we need the following information; A summary of operative procedures to be used for the steam generator secondary water chemistry control and monitoring program, addressing the following:
- 1. Identify the sampling schedule for the critical chemical and other parameters and the control points or limits for these parameters for each operating mode of the plant, i.e., dry lay-up, cold shutdown, hot standby /s'hutdown, and power operation.
- 2. Identify the procedures used to measure the values of the critical parameters, i.e.,' standard identifiable procedures and/or instruments.
- 3. Identify the sampling points, considering as a minimum the steam generator blowdown, the hot well discharge, the fe'edwater, and the demineralizer effluent. We recommend a process flow chart similar to that in EPRI NP-2704-SR "PWR Secondary Water Chemistry Guidelines."
- 4. State the procedures for recording and management of data, defining corrective actions for various out-of-specification parameters.
- 5. Identify (a) the authority responsible for interpreting the data and initiating action and (b) the sequence and timing of administrative events required to initiate corrective action.
10.4.6 Condensate Cleanup System I,NTRODUCTION 1
The purpose of the Condensate Cleanup System is to remove dissolved and i
suspended solids from the condensate in order to maintain a high quality
r'\
of the feedwater being supplied to the steam generators under all normal plant conditions (startup, shutdown, hot standby, power operation). This is '
accomplished by directing the full flow of condensate to a set of mixed bed ;
~~
domineralizer units. Since the demineralizers need periodic resin ;
regene*ation, spare units are provided in the system to replace units taken :
out of service. The system provides final polishing of the secondary cycle condensate water.
EVALUATION The condensate cleanup system is designed to assist in the control of the :
secondary side water chemistry and is part of the total control system.
4 The condensate cleanup system includes all components and equipment necessary for the removal of dis' solved and suspended impurities which may be present
~
in the condensate.
1 We have reviewed the CCS equipment design, daterials and system operation in accordance with Section 10.4.6 of Standard Review Plan, NUREG-0800.
The system meets the requirements .for condensate cleanup capacity, provides ,
effluent of the required purity, and contains adequate instrumentation to monitor the effectiveness of the system. The system is connected to radio- :
active waste disposal systems to allow disposal of spent resin or regenerant ;
solutions when required. We have reviewed the sampling equipment, sampling l locations, and instrumentation to monitor and control the CCS process [
parameters. On the basis of this review, we find that the instrumentation and sampling equipment provided is adequate to monitor and control process j parameters. '
l Based on our review of the applicant's proposed design criteria and design bases for the condensate cleanup system and the requirements for operation
- l. .. of the system, we conclude that the design of the condensate cleanup system l and supporting systems meets our guidelines and is, therefore, acceptable.
The secondary water chemistry monitoring and control program is evaluated
- i in Section 10.3.5. ,
,~; /
CONCULSION Based on the foregoing evaluation, we. conclude that the condensate cleanup system meets our guidelines, and is therefore, acceptable.
10.4.8 Steam Generator Blowdown System INTRODUCTICN The Steam Generator Blowdown System (BDS) is designed to traintain the specified water chemistry in the steam generators during all operational modes. The system continuously removes particulate impurities from the bicwdown flow by directing the flow through electromagnetic filters and demineralizers before returning to the condenser.
EVALUATION The steam generator blowdown system (SGBS) controls the concentration of chemical impurities ana radioactive materials in the secondary coolant. The scope of review of the SGBS included piping and instrumentation diagrams, seismic and quality group classifications, design process parameters, and instrumentation and process controls. The review has included the applicant's evaluation of the proposed system operation and the applicant's estimate of the controlling process parameters.
The steam generator blowdown system is monitored continuously for radiation in the cecondary side of the steam generator. Radioactive blowdown is handled routinely in the domineralizer system and the electromagnetic filters.
Backwash fluids are handled in Secondary Particulate Waste System and the Radwaste System.
The Steam Generator Blowdown System from steam generator nozzles to the BEX area is designed to seismic Category 1 and ASME-III Class 2 requirenents up to last isolation valve and downstream up to BEX area. The Steam Generator Blowdown
T System from BEX area to Seismic Interface Restraint System is designed to seismic Category 1 and ANSI B31.1 requirements. Thus, the SGBS meets the quality standards requirements of General Design Criterion 1 and the seismic requirements of General Design Criterion 2.
The secondary water chemistry monitoring and control program is evaluated in Section 10.3.5.
We have reviewed the SGBDS in accordance with Section 10.4.8 of Standard Review Plan, NUREG-0800.
Instrumentation and automatic controls are provided to monitor and control the operation of the blowdown system, with provision for sampling of the blowdown, in conformance with the guidelines of Branch Technical Position MTEB 5-3. ,'
CONCLUSION Based on the foregoing evaluation, we conclude that the proposed steam generator blowdown system meets our guidelines and is, therefore, acceptable.
4 9
i l .