ML20136F278
Text
d~b,.
"g
~
UNITED STATES
?
NUCLEAR REGULATORY COMMISSION o
{,l4 WASHINGTON, D. C. 20555 j
-4
- Q p
SE? 2 S icg Docket Nos. 50-424/425 MEMORANDUM FOR: ME;Mtihf tiihictor for Licensing Division of Licensing FROM:
William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering
SUBJECT:
ORAFT SAFETY EVALUATION REPORT, V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 & 2, GEORGIA POWER COMPANY, Plant Name:
Vogtle Electric Generating Plant, Unit 1 & 2 Suppliers:
Westinghouse; Bechtel Licensing Stage:
OL Decket Nos.:
50-424/425 Responsible Branch and Project Manager:
LB #4, M. Miller Reviewers:
B. Turovlin, P. Wu Description of Task: Operating License Review O
Status:
Draft SER Complete (3 open items)
The Chemical Engineering Branch has reviewed FSAR Sections 5.2.3, 5.4.2.1,
~
6.1.1, 6.1.2, 9.1.2, 9.1.3, 9.3.2, 9.3.4, 10.3.5, 10.4.6 and 10.4.8 through Amendment 8 against the criteria of NUREG-0800, (Standard Review Plan) and Item II.B.3 of NUREG-0737.
Our draft Safety evaluation Report is' enclosed.
We found Sections 6.1.1, 6.1.2, 9.1.2, 9.1.3, 9.3.2A, 9.3.4, 10.4.6, and 10.4.8 acceptable.
We need additional information to complete our evaluation of Sections 9.3.2B (Post-Accident Sampling System) and 10.3.5 (Secondary WaterChemistry).
We have also reviewed the materials compatibility aspects of Sections 5.2.3, 5.4.2.1 and 6.1.1.
The results of this review have been transmitted to MTEB on January 3, 1984.
To meet the schedule for the SER, the applicant should provide the requested information by February 15, 1985.
Contact:
B. Turov11n P. Wu x28556 x28555
's"
~
f%f09% e c27g.
Y
5.. '
..- r ~
rt.
SE? 2 8 1Ses-Thomas M. Novak -"
Our SALP input is enclosed.
., U, t i c..., i, r- ; -.; '.,
t s,
William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology i
Division of Engineering
Enclosures:
As stated cc:
R. Vollmer D. Eisenhut c"-
V. Benaroya C. McCracken E. Adensam M. Miller S. Pawlicki T. Sullivan J. Weeks (BNL)
B. Turov11n P. Wu D. Smith gw,
J O
p as Lt
9 Draft Safety Evaluation Report By the Office of Nuclear Reactor Regulation Related to Operation of Vogtle Electric Generating Plant Units 1 & 2 Georgia Power Company Docket Nos. 50-424/425
.g g
Post-Accident Emeraency Coolina Water Chemistry 6.1.1 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.
During the containment spray injection phase, the applicant will educt 30 weight percent sodium hydroxide into the containment spray solutfor. which
,y is supplied from the refueling water storage tank at a concentration of 2000 150 ppm baron.
During the containment spray recirculation phase a final pH of 7.0 to 9.0 will be achieved in the sump once the borated water has thoroughly mixed with the educted sodium hydroxide.
Evaluation We evaluated the pH of the containment sump water following mixture in the containment sump with the educted sodium hydroxide. We verified by independent calculations'that sufficient sodium hydroxide is available l
to raise the containment sump water pH to between 7.0 and 9.0.
This is consistent with the minimum pH of > 7.0 required by BTP-MTEB 6-1 to reduce the probability of stress-corrosion cracking of austenitic stainless steel components. We will include in the Technical Specifications surveillance
)
requirements to verify that the 30 weight percent sodium hydroxide does not
~
deteriorate.
i 4
f t
11
- ~
III. CONCLUSION On the basis of the above evaluation, we conclude that the post-accident emergency core cooling water chemistry meets the requirements of SRP Section 6.1.1, BTP-MTEB 6-1 and GDC 14 and is therefore acceptable.
6.1.2 Organic Materials I. Introduction This evaluation is conducted to verify that protective coatings applied inside containment meet the testing requirements of ANSI N101.2 (1972) and the quality assurance of Regulatory Guide 1.54.
Compliance with these requirements provides assurance that the protective coatings will not fail under DBA conditions and generate e
f significant quantities of solid debris or combustible gas which could complicate the accident conditions.
In the FSAR the applicant states that paints and protective coatings applied to exposed surfaces will be applied in accordance with the quality assurance requirements of Regulatory Guide 1.54.
Additionally, the applicant stated that protective coating which are applied in the containment will meet the requirements of ANSI N101.2 (1972) under simulated DBA conditions.
II.
EVALUATION Based on the applicants compliance with the applicable Regulatory Guide and ANSI Standard, we conclude that the protective coating systems and their applications are acceptable and meet the
~
.Y l
m '
requirements of Appendix B to 10 CFR Part 50.
This conclusion is based on the applicant, having met the quality assurance requirements of Appendix B to 10 CFR Part 50 since the coating systems and their applications. ineet the positions of Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to Water-cooled Nuclear Power Plants" and the requirennts of ANSI N101.2, " Protective Coatings (Paints) for
]
Light Wa~ter Nuclear Reactor Containment Facilities." These measures demonstrate their suitability to withstand a postulated design basis accident-(DBA) environment.
The control of combustible gases that can potentially be generated from the organic materials and from qualified and unqualified paints is reviewed under Section 6.2.5.
The consequences of soild debris that can potentially be formed fro'n unqualified paints are reviewed under Section 6.2.2.
, &q 1
CONCLUSION On the basis of the above evaluation, we conclude that the organic materials meet the testing requirements of ANSI N101. 2 and the positionsofRegulatoryguide1.54andare,therefoie, acceptable.
9.1.2 Spent Fuel Storage 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
- assemblies) wetted by the pool water, in accordance
'with Section 9 1 2 of Standard Review Plan NUREG-0800 and " Review and Acceptance of Spent Fuel Storage and Handling Application, April 1978".
n s
i 4
e b
'-tw+-
o--
4 m-._=,+-
,,.w-*e e
isv
-A--
w
--~c-9
--m
-+-M i
reW-w-
=- w W-T--
4-*r**
l.
. The spent fuel racks will be constructed of Type 304 stainless steel.
The spent fuel pool liner is constructed of stainless steel.
The spent fuel storage rack configuration is composed of individual storage cells inter-connected to form an integral structure.
The pool contains oxygen-saturated demineralized water containing boric acid. The water chemistry control of the spent fuel pool has been reviewed elsewhere and found to meet NRC recommendations.
EVALUATION The pool liner, rack lattice structure and fuel storage tubes are stainless steel which is compatible with the storage pool environment.
In this environment of oxygen-saturated borated water, the corrosive deterioration
-5 of the Type 304 stainless steel should not exceed a depth of 6.00 X 10 inches in 100 years, which is negligible relative to the initial thickness.
,[
Dissimilar metal contact corrosion (galvanic attack) between the stainless steel of the pool liner, rack lattice structure, fuel storage tubes, and the Inconel and the Zircaloy in the spent fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are therefore at similar potentials.
We find that the corrosion that will occur in the spent fuel storage environ-ment should be of little significance during the 40 year life of the plant.
Components in the spent fuel storage po.o1 are constructed of alloys which have a low differential galvanic potential between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosion.
We also find that the environmental compatibility and stability of the materials used in the spent fuel storage pool is adequate based on test data and actual service experience in operating reactors.
p
o.
CONCLUSION Based on the above evaluation, we conclude that the selection of appropriate materials of construction by the applicant meets the requirements of 10 CFR Part 50, Appendix A, Criterion 61, having a capability to permit appropriate periodic inspection and testing of components, and Criterion 62, preventing criticality by maintaining structural integrity of components and of the boron poison and is, therefore, acceptable.
9.1.3 Fuel Pool Cleanup System I. INTRODUCTION The spent fuel pool cleanup system is designed to remove radioactive species and other impurities for the purpose of maintaining area radiation levels as low as is reasonably p,'
achievable during fuel handling and other maintenance operations and to maintain water clarity for the pupose of facilitating fuel bundle movement.
The fuel pool cleanup system services the spent fuel pool, refueling pool and fuel transfer canal.
The system consists of two full design flow purification loops and one surface skimmer.
The fuel pool cleanup system will be used continuously during refueling operations to reduce radioactivity and maintain clarity of the refueling pool.
Subsequent to refueling, the cleanup system will be manually operated intermittently to maintain clarity and reduce radioactivity of the spent fuel pool area to < 2.5 MREM /Hr.
Manual imtermittent use of the fuel pool cleanup system will usually be initiated concurrent with the fuel pool cooling system but can be initiated by
~
the need for spent fuel pool clarity or due to high f'.
radiation (>2.5MilEM/HR)inthearea.
k
. -3 During operation of the fuel pool cleanup system, samples will be taken for chemical analysis of baron, chloride, fluoride, pH and radioactivity on an intermittent basis.
Additionally, the system decontamination factor and differential pressure for the filters and ion exchangers will also be monitored to determine the need for filter or resin replacement.
The specific sampling and monitoring frequency will be based on the frequency which the system is in use.
II. EVALUTION We have determined that the spent fuel pool cleanup system (1) provides the capability and capacity of removing radioactive materials, corrosion products and impurities from the pool water, and thus meets the require-ments of General Design Criterion 61 in Appendix A to 10 CFR Part 50, as g.g it relates to appropriate filtering systems for fuel storage; (2) is t'
l capable of reducing occupational exposure to radiation by removing radio-active products from the pool water, and thus meets the requirements of Section 20.l(c) of 10 CFR Part 20, as it relates to maintaining radiation exposures as low as is reasonably achievable; (3) confines radioactive materials in the pool water into the demineralizer and filters, ani thus meets Regulatory Position C.2.f(2) of Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable," as it relates to reducing the spread of contaminants from the source; and (4) removes suspended impurities from the pool water by filters, and thus meets j
Regulatory Position C.2.f(3) of Regulatory Guide 8.8, as it relates to removing crud through physical actions.
III. CONCLUSION On the basis of the above evaluation, we conclude that the spent fuel pool cleanup system meets GDC 61, Section 20.l(c) of 10 CFR Part 20 lI and the appropriate sections of Regulatory Guide 8.8 and, therefore, is acceptable.
m
- .-s 4
9.3.2A process Sampling Systems I. INTRODUCTION Process sampling is accomplished by a primary sampling system and a secondary sampling system.
The primary sampling system is designed to collect water and gaseous samples contained in the reactor coolant system and associated auxiliary system process streams during all normal modes of operation.
The secondary sampling system is designed to collect water and steam from the secondary cycle.
Continuous secondary samples are analyzed automatically for pH, conductivity, dissolved oxygen, residual hydrazine and sodium.
Additionally, grab samples are obtained for confirmatory analysis and other chemical species.
Provisions are made to assure that representative samples are obtained from well mixed streams or volumes of effluent by the
)
selection of proper sampling equipment and location of sampling points as well as proper sampling procedures.
The primary sample lines penetrating the containment are each equipped with two normally closed pneumatically operated isolation valves, which if open, close on a containment isolation actuation signal.
II.
EVALUATION Our review included the provisions to sample all principal fluid process streams associated with plant operation and the applicant's design of these systems, including the location of sampling points, as shown on piping and instrumentation diagrams.
4
-~-,
v s.
We determined that the process sampling system meets (1) the requirements of General Design Criterion 13 in Appendix A to 10 CFR Part 50, by sampling the reactor coolant, the safety infection tanks, the refueling water storage tank, the boric acid mix tank, and the boron injection tank for boron concentration, which can affect the fission process for normal operation, anticipated operational occurrences, and accident conditons; (2) the requirements of GDC 14, by sampling the reactor coolant and the secondary coolant for chemical impurities to ensure that the reactor coolant pressure boundary will have a low probability of abnormal leakage, rapidly propagating failure, and gross rupture; (3) the requirements of GDC 26 by sampling the reactor coolant, the refueling water storage tank, and the boric acid mix tank for boron concentrations for controlling the rate of reactivity changes; (4) the requirements of GDC 63 by sampling the spent fuel pool and the gaseous radwaste storage tank for
- (p,'
radioactivity to detect conditions that may result in excessive radiation levels; and (5) the requirements of GDC 64 by sampling the reactor coolant, the pressurizer, the steam generator blowdown, the sump inside containment, the containment atmosphere and the gaseous radwaste storage tank, for radioactivity that may be released from normal operations, including anticipated operational occurrences and from postulated accidents.
We further determined that the 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.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
~. -
em
-9 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 60 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 up to and including the first isolation f]
valves to which each sampling line and component is connected, and-I thus meets the quality standards requirements of GDC 1 and the seismic requirements of GDC 2.
III.
CONCLUSION Based on the above evaluation, we find the proposed process sampling system acceptable.
9.3.28.
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 accept-
"able sampling and analysis system are specified in NUREG-0737, Item II.B.3.
7,.;
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 non-volatile isotopes),
hydrogen in the containment atmosphere and total dissolved gases or hydrogen, boron, and chloride in reactor coolant samples.
EVALUATION Criterion (1):
The applicant shall have the capability to promptly obtain reactor coolant samples and containment.atmsophere samples.
The combined time allotted for sampling and analysis should be three hours or
- y less from the time a decision is made to take a sample.
The applicant has provided in-line sampling and analysis capability to promptly obtain and analyze reactor coolant samples and containment atmosphere samples within three hours from the time a decision is made to take a sample. We find that these provisions meet Criterion (1) and are, therefore, acceptable.
Criterion (2):
ll The licensee 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 that may be indicators of the degree of core
~
damage (e.g., noble gases, lodinas and cisiums, and non-volatile isotopes);
,.s s
[ '
~
E
. b) hydrogen levels in the containment atmosphere; i
c) dissolved gases (e.g...H ), chloride (time allotte.1 for 2
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 the capabi,lity to collect diluted or undiluted liquid and gaseous reactor coolant and containment atmopshere grab samples that can be transported to the onsite radiological and chemical laboratory for hydrogen, oxygen, pH, conductivity, boron, chloride, and radionuclide analyses. Arrangements have been made with an off-site laboratory for backup and supplemental analyses.
We find that these provisions partially
}
meet Criterion (2). The applicant should provide a plant specific core damage estimating procedure which takes into consideration plant parameters such as core exist temperature, water level in reactor vessel, containment radiation monitors, and hydrogen analyses.
Criterion (3):
Reactor coolant and containment atmosphere sampling during post-accident conditions shall not require an isolated auxiliary system (e.g., the letdown system, reactor water cleanup system (RWCUS))
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 PASS valves which are not accessible after an accident have been selected to withstand the
'specified service environment.
These provisions meet Criterion (3) and are, therefore, acceptable.
ye c -. - -
.-m
_,__--_.,--.-3
~
y
,. 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 concentration is recommended, but is not mandatory.
2 The PASS can quantify both dissolved oxygen and dissolved hydrogen with unpressurized reactor coolant samples using a hydrogen gas chromatograph and an Orbisphere oxygen analyzer.
Dissolved oxygen can be measured to less than 0.1 ppm.
We have determined that these provisions meet Criterion (4) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable.
Criterion (5):
f' Q 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 primary containment systems and tne 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 being 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 cooling water is neither seawater nor brackish, and more than one barrier is provided between primary containment systems and the cooling water; therefore, chloride analysis wil be performed within 4 days. We have determined that these provisions meet Criterion (5) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable.
(
J
~
e-
~,,, - - -
-w.
~
,--e,
---e v------w-
Criterion (6):
The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that 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) (l.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 FA (NUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H. R. Denton to all licensees).)
The PASS is designed to permit in-line analysis of the reactor coolant, emergency sumps, and containment atmosphere in accordance with the exposure criteria of GDC 19.
Remote in-line sampling is performed from the radiochemistry laboratory.
In addition, the PASS has the capability to p
obtain both diluted and undiluted backup grab samples. We have determined
' ],
that these provisions meet Criterion (6) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable.
Criterion (7):
.The analysis of primary coolant samples for boron is required for PWRs.
The PASS has the capability to perform boron analysis in primary coolant samples.
We have determined that these provisions meet Criterion (7) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable.
Criterion (8):
If in-line monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing the samples.
Established planning for analysis at offsite facilities
., is acceptable.
Equipment provided for backup sampling shall be capable of providing at least one sample per week until the accident condition no longer exists.
In-line monitoring is used as the primary method of performing the required analysis.
The PASS has the capability to obtain both diluted and undiluted backup grab samples.
In the event both in-line monitoring and grab sampling analysis capability fail, arrangements have been made to send the samples to Oak Ridge National Laboratories in a licensed shipping cask.
Provisions for in-line monitor flushing have been provided to reduce plateout, crud buildup, and radiation exposure of coo;onents, the panel tubing and monitors are flushed after every panel exercise.
We have determined that these provisions meet criterion (8) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable,
'9, 7
Criterion (9):
The applicant's radiological and chemical sample analysis capability shall include provisions to:
a) Identify and quantify isotopes of the nuclide categories discussed 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 1 pCi/g to 10 Ci/g.
b) Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample
~
analysis will provide results with an acceptably small error f%,
.o-
. 9 (approximately a factor of 2).
This can be accomplished through the use of sufficient shielding around samples and outside sources, and by the use of a ventilation system design which will control the presence of airborne radioactivity.
The radionuclides in both the primary coolant and the containment atmosphere will be identified and quantified.
Provisions are available for diluted reactor coolant samples to minimize personnel exposure.
The PASS can perform radioisotope analyses at the levels corresponding to the source term given in Regulatory Guide 1.4, Rev. 2.
Radiation background levels will be restricted by shielding.
Radiological and chemical analysis facilities are provided to obtain results within an acceptably small error (approximately a factor of 2). We find that these provisions meet Criterion (9) and are, therefore, acceptable.
Criterion (10):
Accuracy, range, and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant systems.
The applicant stated that the FASS is designed to provide accuracy, range, and sensitivity necessary to allow the operator to determine the radiological and chemical status of the sample.
We determined that these provisions do not meet Criterion (10).
The applicant should provide information regarding accuracy, range, and sensitivty of.the PASS instruments and analytical procedures in the post-accident water chemistry and radiation environment consistent with the recommendations of Regulatory Guide 1.97, Rev. 3, and the clarifications of NUREG-0737, Item II.B.3, Post-Accident Sampling Capability, transmitted to the applicant on March 13, 1984.
In addition, the frequency for demonstrating operability of procedures and instrumentation "and retraining of operators on a minimum semi-annular basis should be provided.
/
h
C m '
p Criterion (11):
In the design of the post-accident sampling and analysis capability, consideration should be given to the following items:
a) Provisions 'should be made for purging sample lines, for reducing plateout in sample lines, 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 limit 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 possible 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 p(j system.
b) The ventilation exhaust from the sampling station should be filtered with charcoal adsorbers and high-efficiency particulate air (liEPA) filters.
The applicant has addressed provisions for purging to ensure samples are representative, size of sample line, isolation valves to limit reactor coolant loss from a failure of the sample line, and for ventilation exhaust from PASS to be filtered through charocal adsorbers and HEPA filters.
To limit iodine plateout, the containment air sample line is heat traced.
The post-accident reactor coolant and containment atmosphere samples will be representa-tive of the reactor coolant in the core area and the containment atmosphere.
We determined that these provisions meet Criterion (11) of Item II.B.3 of NUREG-0737, and are, therefore, acceptable.
P j
5'.
~
. 3
~
CONCLUSION We conclude that the post-accident sampling system meets nine of the eleven criteria in Item II.B.3 of NUREG-0737. Additional informtion is needed to complete our review of the remaining two criteria:
(2) provide a plant specific procedure to estimate the extent of core damage.
(10) provide information on the accuracies, sensitivities, and performance of the PASS instrumentation and analytical procedures in the post-accident. water chemistry and radiation environment.
Provide the frequency for demonstrating operability of procedures, and instrumentation and retraining of operators on semi-annual basis.
.s.,.
9.3.4 Chemical and Volume Control System f N.
' _4 I.
INTRODUCTION-The chemical and volume control system (CVCS) is designed to control and maintain reactor coolant inventory and to control the boron concentration in the reactor coolant through the process of charging (makeup) and letdown (drawing off).
The CVCS purifies the primary coolant by passing letdown flow through heat exchangers and purification ion exchangers.
The CVCS is also designed to provide reactor coolant pump seal injection flow and to collect the controlled bleedoff from the seals.
Three charging pumps, one positive displacement and two centrifugal, supply high pressure injection (charging) of borated water into the reactor coolant for normal and emergency boration.
The volume control tank serves as a surge for the reactor coolant system, to provide for control of hydrogen concentration in the reactor coolant, and to provide a reservoir of makeup for the charging pumps.
The boric acid makeup system in conjunction with the boron thermal Tegeneration system, provides for boron additions to compensate for reactivity changes and to provide shutdown margin for maintenance and refueling operations or emergencies.
Therefore, the charging portion of the system is designed to seismic Category I requirements and contains redundant active ccmponants and an alternate flow path in order to meet the single failure criteria.
,t,'..
w 18-II.
EVALUATION The CVCS including boron recovery system, includes components and piping associated with the system from the letdown line of the primary system to the charging lines that provide makeup to the primary system and the reactor coolant pump seal water system.
The basis for acceptance in our review has been conformance of the applicant's design of the CVCS system with the following regulations and regulatory guides:
(1) the requirements of General Design Criterion 1 and the guidelines of Regulatory Guide 1.26 by assigning quality group classifications to system components in accordance with the importance of the safety function to be performed; (2) the requirements of General Design Criterion 2 and the guidelines of Reguatory Guide 1.29 by designing safety-related portions of the system to seismic Catep'n 1 requirements; (3) the requirements of
,e General Design Criterion 14 oy maintaining reactor coolant purity and
{.
material compatibility to reduce corrosion and thus reduce the probability of abnormal leakage, rapid propagating failure, or gross rupture of the reactor coolant pressure boundary; (4) the requirements of General Design Criterion 29 as related to the reliability of the CVCS to provide negative reactivity to the reactor by supplying borated water to the reactor coolant system in the event of anticipatad operational occurrences; and (5) the requirements of General Design Criteria 60 and 61 with respect to confining radioactivity by venting and collecting drainage from the CVCS components through closed systems.
t II.
CONCLUSION Based on the above review, we conclude that the design of the chemical and volume control system and supporting system is acceptable and meets the requirements of General Design Criteria 1, 2, 14, 29, 60 and 61, and is',
'therefore, acceptable.
I y&
..-.m..
n,-..
,4
, -, -,-.~
m,
-,w,-,-
1
~
l -
10.3.5 Secondary Water Chemistry Introduction In late 1975, we incorporated provisions into the Standard Technical 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 con-tain either these provisions or a requirement to establish these pro-visions after baseline chemistry conditions have been determined.
The intent of the provisions was to provide added assurance that the operators of newly licensed plants would properly monitor and control secondary water chemistry to limit corrosion of steam generator com-ponents such as tubes and tube support plates.
M In a number of instances, the plant Technical Specifications have signifi-cantly restricted the operational flexibility of some plants with little or no benefit with regard to limiting degradation of steam generator tube and the tube support plates.
Based on this experience and the knowledge gained in recent years, we have concluded that Technical Specification limits are not the most effective way of assuring that steam generator degradation will be minimized.
Due to the complexity of the corrosion phenomena involved and the state-of-the-art as it exists today, we are of the opinion that, in lieu of specifying limiting conditions in the plant Technical Specification, a more effective approach would be to specify a technical specification that required the implementation of secondary water chemistry monitor-ing and control program containing appropriate procedures and admini-strative controls. This has been the approach for control of secondary
~water programs since 1979.
m
^ ~ '.. -
p..
. ~s The required program and procedures 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 con-ditions in the steam generators.
In our view, plant operation follow-ing such procedures would provide adequate assurance that licensees would devote proper attention to controlling secondary water chemistry, while also providing the needed flexibility to allow them to deal effectively with an off-normal condition that might arise.
EVALUATION The applicant did not provide details of a secondary water chemistry monitoring and control program.
The applicant stated in Amendment 5 that the requested information would be provided at a later date.
To complete our review, we need the following information:
'\\
\\
j A summary of operative procedures to be used for the steam generator s,
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 / shutdown, 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 dischcge, the feedwater, and the demineralizer effluent.
We recommend a process flow chart similar to that in EPRI NP-2704-SR, "PWR Secondary Water Chemistry Guidelines."
N
- s
. 4.
State the procedures for recording and management of data, defining corrective actions for various out-of-specification parameters.
The procedures should define the allowable time for correction of out-of-specification parameters. We recommend multiple levels of time allowable for providing correction based upon the amount of out-of-specification of the variable.
(See EPRI NP-2704-AR above).
Because of the significance of condenser in-leakage the chemistry program should include a corrective action provision such that a condenser inservice inspection program will be initiated if condenser leakage is of such a magnitude that power reduction is required (action level 2 of the EPRI/SGOG guidelines) more than once per three month period.
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 cor'rective action.
6 10.4.6 Condensate Cleanup System Introduction The purpose of the condensate cleanup system is to remove dissolved and suspended solids from the condensate in order to maintain a high quality of the feedwater being supplied to the steam generators under all nornal plant conditions (startup, shutdown, hot standby, power operation).
This is accomplished by directing the full flow of condensate to a set of filter-mixed bed demineralizer units.
Since the demineralizers need periodic resin regeneration, 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.
-22 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.
The condensate cleanup system includes all components and equipment necessary for the removal of dissolved and suspended impurities which may be present in the condensate.
We have reviewed the CCS equipment design, materials 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, and contains adequate instrumentation to monitor the effectiveness of the system.
We have reviewed the sampling equipment, sampling locations, and instrumenta-
[
tion to monitor and control the CCS process parameters.
On the basis of this review, we find tht the instrumentation and sampling equipment provided is adequate to monitor and control process parameters.
Based on our review of the applicant's criteria and design bases for the l
condensate cleanup system and the requirements for cperation of the system, l
we conclude that the design of the condensate cleanup system and supporting l
systems meets our guidelines and is, therefore, acceptable.
The secondary water chemistry monitoring and control program is evaluated in Section 10.3.5.
i l
J
j.. ";..,
O a Conclusion Based on our review, we conicude that the design of the condensate cleanup system supporting systems is acceptable and meets the primary boundary integrity requirements of General Design Criterion 14.
This conclusion is based on the applicant having met the requirements of GDC 14 as it relates to maintain-ing acceptable chemistry control for PWR secondary coolant during normal operation and anticipated operational occurrences by reducing corrosion of PWR steam generator tubes and materials, thereby reducing the likelihood and magnitude of reactor piping failures and of primary-to-secondary coolant leakage.
This requirement has been met by the applicant's design of the CCS meeting the Branch Technical Position MTEB 5-3 for PWRs.
Based on the foregoing, we conclude that the condensate cle'anup system meets our guidelines and is, therefore, acceptable.
$D 10.4.8 Steam Generator Blowdown System Introduction The steam generator blowdown system (SG85) is used in conjunction with the condensate demineralizer, chemical addition, and sample systems to control the chemical ccmposition of the steam generator shell side water within specified limits during all operating modes.
The hlowdown fluids are directed to a set of heat exchar.gers to cool the fluiis and then to a series of filter-mixed bed demineraliz,er, the blowdown tank and then to the condenser.
Evaluation The SGBS controls the concentration of chemical impurities and radioactive
' materials in the secondary coolant.
The scope of review of the SGBS included piping and instrumentation diagrams, seismic a M quality group classifications, design process parameters, and instrumentation and process controls.
n 3:v;;
i The portion of the steam generator blowdown system up to and including the containment isolation valves is Seismic Categroy I and designated
.ASME II Class 2.
All other piping and equipment in the steam generator blowdown system is not safety related and is designed and built to ANSI B31.1 requirements.
Thus, the SGBS meets the quality standards requirements of General Design Criterion 1 and seismic requirements of General Design Criterion 2.
Instrumentation and automatic controls are provided to monitor and con-trol 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
- fyrs, Based on our evaluation, the SGBS meets the primary boundary material lj-integrity requirements of General Design Criterion 14 as it relates to maintaining acceptable secondary water chemistry control during normal operation and anticipated operational occurrences by reducing corrosion of steam generator tubes and materials thereby reducing the likelihood and magnitude of primary-to-secondary coolant laakage.
Based on the foregoing evaluation, we conclude that the proposed steam generator blowdown system meets our guidelines and, is therefore, acceptable.
Summary Open item:
A) 9.3.28 Post-Accident Sampling System (NUREG-0737, II.B.3)
(2) provide plant specific procedures to estimate core damage.
(10) provide information on accuracy, sensitivity, and performance of PASS instrumentation.and procedures.
c.
- p T,.-.
3.'
, 8) 10.3.5 Secondary Water Chemistry 1)
Sampling schedule and control limits 2)
Procaudres 3)
Sampling points 4)
Data managment 5)
Responsible authority 1
$~
~,
}
I b
i e
5 i
e-