ML20136B812
| ML20136B812 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Vogtle |
| Issue date: | 10/04/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML082840446 | List:
|
| References | |
| FOIA-84-663 NUDOCS 8601030018 | |
| Download: ML20136B812 (16) | |
Text
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i ATTACHMENT l
SAFETY EVALUATION REPORT V0GTLE UNITS 1 & 2 DOCKET NO. 50-424/425 MATERIALS ENGINEERING BRANCH COMPONENT INTEGRITY SECTION Turbine Missiles Introduction During the past several years the results of turbine inspections at operating nuclear facilities indicate that cracking to various degrees has occurred at the inner radius of turbine disks, particularly those of Westinghouse design.
Within this time period, there has actually been a Westinghouse turbine disk failure at one facility owned by the Yankee Atomic Electric Company.
Furthermore, recent inspections of General Electric turbines have also resulted in the discovery of disk keyway cracks.
Stress corrosion has been identified by both manu-facturers as the operative cracking mechanism The staff has followed these developments closely.
Our primary safety objective is the prevention of unacceptable doses to the public from releases of radioactive contaminants due to damage to plant safety-related structures, systems, and components due to missile generating' turbine
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failures.
Based on previous staff reviews and various estimates by l
oth'ers (Refs.1 and 2) for a variety of plant layouts the staff concludes that "if a turbine missile is generated" the probability of l
unacceptable damage to safety-related structures, systems, and components 3
.2 is in the neighborhood of 10 or 10 per year depending on whether the turbine orientation is favorable or unfavorable.
In view of this and operating experience, we have shifted the review emphasis to the l
prevention of missile generating turbine failures.
In keeping with this shift of emphasis the staff has recently set turbine missile i
l generation probability guidelines for determining (a) turbine disk l
ultrasonic inservice inspection frequencies, and (b) turbine control and overspeed protection systems maintenance and testing schedules.
my 8601030018 851127 PDR FOIA BELL 84-663 PDR
-- +
. Criteria that Must be Met to Demonstrate Compliance With the Regulation According to General Design Criterion 4 of Appendix A to 10 CFR Part 50, nuclear power plant structures, systems, and componerts important to safety shall be appropriately protected against the dynamic effect of missiles.
Failures of large steam turbines of the main turbine generator have the potential for ejecting large high energy missiles that can damage plant structures, systems and components. The overall safety objective of the staff is to assure that structures, systems, and components important to safety are adequately protected from potential turbine missiles.
Of those systems important to safety, this topic is primarily concerned with safety-related systems; i.e.,
those structures, systems, or components necessary to perform required safety functions and to ensure:
1.
The integrity of the reactor coolant pressure boundary,
, -7 y.,
2.
The capability to shut down the reactor and maintain it in
.a safe shutdown condition, or 3.
The capability to prevent accidents that could result in potential offsite exposures that are a significant fraction of the guideline exposures of 10 CFR Part 100, " Reactor Site Criteria."
Typical safety-related systems are listed in Regulatory Guide (RG) 1.117.
The probability of unacceptable damage due to turbine missiles (P ) is 4
generally expressed as the product of (a) the probability of turbine failure resulting in the ejection of turbine disk (or internal structure)
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fragments through the turbine casing (P ), (b) the probability of ejected j
missiles perforating intervening barriers and striking safety related structures, systems, or components (P ), and (c) the probability of 2
struck structures, systems, or components failing to perform their
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safety function (P )*
3 According to NRC guidelines stated in Section 2.2.3 of the Standard Review Plan (SRP) NUREG-0800,and RG 1.115, the probability of unacceptable damage from turbine missiles should be less than or equal to about one chance in ten million per year for an individual
.7 plant, i.e., P 5 10 per year.
4 Past Procedure for Demonstrating Compliance with Regulations In the past, analyses for construction permit (CP) and operating license (OL) reviews assumed the probability of missile generation (P ) to be 1
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.4 approximately 10 per turbine year, based on the historical failure rate (Ref.1).
The strike probability (P ) was estimated (Ref. 3) 2 based on postulated missile sizes, shapes, and energies, and on avail-able plant specific information such as turbine placement and orientation, number and type of intervening barriers, target geometry, and potential missile trajectories.
The damage probability (P ) was generally assumed 3
to be 1.0.
The overall probability of unacceptable damage to safety related systems (P ), which is the sum over all targets of the product 4
of these probabilities, was then evaluated for compliance with the NRC safety objective.
This logic places the regulatory emphasis on the strike probability; i.e, having established an individual plant safety
.7 objective of about 10 per year, or less, for the probability of unacceptable damage to safety related systems due.to turbine missiles, 4
.3 this procedure requires that P be less than or equal to 10 2
i "It is well known if that nuclear turbine disk cracks (Refs. 4 and 5) and that turbine stop and control valves fail (Refs. 6 and 7), then disk ruptures can result in the generation of high-energy missiles s
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(Ref. 8).
Furthermore, analyses (Refs. 7 and 9) clearly demonstrate the large effects of inservice testing and inspection frequencies on missile generation probabilities (P ).
It is the staffs view that 1
sufficiently frequent turbine testing and inspection are the best means of assuring that the criteria on the probability, P, of unacceptable damage 4
to safety related structuras, systems, and components are met.
Therefore, it is prudent for turbine manufacturers to perform, and the NRC to review, analyses of turbine reliability, which' include known and likely failure mechanisms, expressed a8 a function of time (i.e., inservice inspection or test intervals).
While the calculation of strike probability is not difficult in princi-ple, for the most part reducing to a straightforward ballistic analysis, it presents a problem in practice.
The problem stems from the fact that numerous modeling approximations and simplifying assumptions are required
_s to make tractable the incorporation of acceptable models based on available data, and (a) properties of missiles, (b) interactions of missiles with barriers and obstacles, (c) trajectories of missiles as they interact with and perforate (or are deflected by) barriers, and (d) identification and location of safety-related targets.
The particular approximations and assumptions made tend to have a large effect on the resulting value of P.
Similarly, a reasonably accurate specification of the damage 2
probability (P ) is not a simple matter due to the difficulty of 3
defining the missile impact energy required to render given safety-related systems unavailable to perform their safety function, and the difficulty of postulating sequences of events that would follow a missile producing turbine failure.
New Procedure for Demonstrating Compliance with Regulations Jhe new approach places on the applicant the responsibility for demon-strating and maintaining a NRC specified turbine reliability by appropriate inservice inspection and testing throughout plant life.
, This shift of emphasis necessitates that the applicant show capability to have volumetric (ultrasonic) examinations performed which are suitable for inservice inspection of turbine disks and shaft, and to provide reports for NRC review and approval which describe their methods for determining turbine missile generation probabilities.
Westinghouse and General Electric have prepared reports for NRC review and approval which describe methods for determining turbine missile generation probabilities for their respective turbines.
The design speed missile generation probability is related to disk design parameters, material properties, and the inservice volumetric (ultrasonic) disk inspection interval (for example, see Ref. 9).
The destructive overspeed missile generation probability is related to the turbine governor and overspeed protection system's speed sensing and tripping characteristics, the design and arrangement of main steam control and stop valves and the q
reheat steam intercept and stop valves, and the inservice testing and inspection intervals for systems components and valves (for example, see Ref. 7).
The manufacturer will provide applicants and licensees with tables of missile generation probabilities versus time (inservice volumetric disk inspection interval for design speed failure, and
-inservice valve testing interval for destructive overspeed failure) for their particular turbine, which will be used to establish inspection and test schedules which will meet NRC safety objectives.
-Due to the uncertainties involved in calculating P, the staff 2
coocludes that P analyses are " ball park" or " order of magnitude" 2
type calculations only.
Based on simple estimates for a variety of plant layo'its (for examples, see Refs.1 and 2), the staff further concludes that the strike and damage probability product can be reasonably taken to fall in a characteristic narrow range which is dependent on the gross features of turbine generator orientation;
'(a) for favorably oriented turbine generators P P tend to lie in 2 3 the range 10' to 10
, and (b) for unfavorably oriented turbine
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.2 generators P P tend to lie in the range 10 to 10 For these 2 3 u
- reasons (and due to weak data, controversial assumptions, and modeling difficulties), in the evaluation of P, the staff gives credit for the 4
product of the strike and damage probabilities of 10 for a favorably
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oriented turbine and 10 for an unfavorably oriented turbine, and does not encourage calculations of them.
These values represent our opinion of where P P lie based on calculations we have done and the results of 2 3 calculations done by others.
It is the staff's view that the NRC safety objective with regard to turbine missiles is best expressed in terms of two sets of criteria applied to the missile generation probability (see Table 1).
One set of criteria is to be applied to favorably oriented turbines, and the other is to be applied to unfavorably oriented turbines.
Applicants andlicensees,withturbinesfrommanufacturerswhohavehadreporps describing their methods and procedures for calculating turbine missile generation probabilities reviewed and accepted by the NRC, are expected to meet the set of criter,ia appropriate to their turbine orientation, as shown in Table 1.
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i TABLE 1 RELIABILITY CRITERIA
-1 Probability, yr Favorably Unfavorably Required Oriented Oriented
_ Licensee Action 4
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-5 A.
P1 < 10 Py < 10 This is the general, minimum reliability requirement ftr loading the turbine and bringing the system on line, B.
10~4 < Py < 10 10-5 < P < 10 If during operation this condition is reached, the turbine
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may be kept in service until the next scheduled outage, at
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vhich time the licensee is to take actkn to reduce P to t
i meet the appropriate A criterion (above) before returning the turbine to service.
C.
10-3 < Py < 10 10-4 < Py < 10 If during operation this condition is reached, the turbine
-2
-3 is to be isolated from the steam supply within 60 days, at j
which time the licensee is to take action to reduce P to t
meet the appropriate A criterion (above) before returning the turbine to service.
D.
10-2 < p 10-3 < P If at any time during operation this condition is reached, 1
the turbine is to be isolated from the steam supply within 6 days, at which time the licensee is to take action to reduce P to meet the appropriate A criterion (above) 2 before returning the turbine to service.
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. Alternative Procedure for Demonstrating Compliance with Regulations Applicants and licensees, with turbines from manufacturers who have not yet submitted reports tn the NRC describing their methods and procedures for calculating turbine missile generation probabilities or who have submitted reports which are still being reviewed by the NRC, are expected to meet the following alternative criteria, regardless of turbine orientation:
A.
The inservice inspection program employed for the steam turbine rotor assembly is to provide assurance that disk flaws that might lead to brittle failure of a disc at speeds up to rksign speed will be detected.
The turbine rotor design should te such as to facilitate inservice inspection of all high stress regions, including disk bores and keyways, without the need for removing n
the disks from the shaft.
The volumetric inservice inspection interval for the steam turbine rotor assembly is to he estab-lished according to the following guidelines:
1.
The initial inspection of a new rotor or disk should be performed before any postulated crack is calculated to grow to more than 1/2 the critical crack depth.
If the calculated inspection interval is less than the scheduled first fuel cycle, the licensee should seek the manu-facturer's guidance on delaying the inspection until the refueling outage.
If the calculated inspection interval is longer than the first fuel cycle, the licensee should seek the manufacturer's guidance for scheduling the first inspection at a later refueling outage.
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Disks that have been previously inspected and found to be free of cracks or that have been repaired to eliminate all indications should be reinspected using the same criterion as for new discs, as described in (1), calcu-lating crack growth from the time of the last inspection.
3.
Disks operating with known and measured cracks should be reinspected before 1/2 the time calculated for any crack to grow to 1/2 the critical crack depth.
The guidance described in (1) should be used to set the j
inspection date based on the calculated inspection i
interval.
4.
Under no circumstances is the volumetric inservice inspection interval for LP disks to exceed approxi-mately 3 years or 2 fuel cycles.
Ins'pections during these refueling or maintenance shutdowns should consist of visual, surface, and volumetric examinations, according to the manufacturer's procedures, of all normally inaccessible parts such as couplings, coupling bolts, LP turbine shafts, blades, and disks, and HP rotors.
Shafts and disks with cracks of depth near to or greater than 1/2 the critical crack depth are to be repaired or replaced.
All cracked couplings and coupling bolts should be replaced.
B.
The inservice inspection and test program employed for the governor and overspeed protection system should provide assurance that flaws or component failures in the overspeed sensing and tripping subsystems, in the main steam control 4
s.:
a.,.
, and stop valves, reheat steam intercept and stop valves, or extraction steam non-return valves that might lead to an overspeed condition above the design overspeed will be detected.
The inservice inspoction program for governor and overspeed protection systems operability should include, as a minimum, the following provisions:
1.
For typical turbine governor and overspeed protection systems, at approximately 3 year intervals, during refueling or maintenance shutdowns, at least one main steam control valve, one main steam stop valve, one reheat intercept valve and one reheat stop valve, and one of each type of steam extraction valves are to be dismantled and visual and, surface examinations conducted "of valve _ seats, discs, and stems.
Valve bushings should 7 ~s 5e inspected and cleaned, and bore diameters should be checked for proper clearance.
If any valve is shown to have hazardous flaws or excessive corrosion or improper clearances, the valve is to be repaire.d or replaced and all other valves of that type dismantled and inspected.
2.
Main steam control and stop valves, reheat intercept and l
stop valves, and steam extraction non-return valves are to be exercised at least once a week during normal operation by closing each valve and observing directly l
the valve motion as it moves smoothly to a fully closed j
position.
3.
At least once a month during normal operation each com-E partment of the electrohydraulic governor system (which modulates control and intercept valves), and the t-I I
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, mechanical overspeed trip mechanism and backup electrical overspeed trip (both of which trip the main steam control and stop valves, and reheat intercept and stop valves) are to be tested.
On-line test failures of any one of these subsystems require repair or replacement of failed components within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the turbine is to be isolated from the steam supply until repairs are completed.
Evaluation The GE turbine generator is intended for baseload operation and also has load following characteristics consistent with the requirements of the Westinghouse nuclear steam supply system.
The placement and orientation of the turbine generator is unfavorable with respect to the station rea'ctor buildings; that is, there are safety related 7s targets inside the low trajectory missile strike zone.
The turbine
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is a tandem-compound, 6-flow, reheat unit with 38-in. last-stage buckets (TC 6F 38-in. LSB).
The turbine unit consists of one double-flow high pressure turbine, three double-flow low pressure turbines, and a rated rotational speed of 1800 rpm.
Destructive Overspeed Failure Prevention The turbine generator has a turbine control and overspeed protection system which is designed to control turbine action under all normal or abnormal conditions and to ensure that a turbine trip from full load will not cause the turbine to overspeed beyond acceptable limits so as to minimize the probability of generating turbine missiles, in accordance with requirements of GDC 4.
The turbine control and overspeed protection system is, therefore, essential to the overall safe operation of the plant.
. Turbine control is accomplished with an electrohydraulic control (EHC) system.
The EHC system consists of an electronic governor using solid state control techniques in combination with a high pressure hydraulic actuating system.
The system includes electrical control circuits for steam pressure control, speed control, load control, and steam control valve positioning.
There are three methoc's of turbine overspeed control protection:
the normal speed governor (EHC), the mechanical overspeed trip mechanism, and the electrical overspeed trip.
The EHC modulates the turbine con-trol valves to maintain desired speed load characteritics.
At 103 percent of rated speed, the EHC will close the governor and intercept valves.
The mechanical overspeed sensor trips the turbine stop, con-trol, and combined intermediate valves by deenergizing the hydraulic fluid system when 110.0 percent of rated speed is reached.
The s,
electrical backup overspeed sensor trips these same valves when 112.0 percent of rated speed is reached by independently deenergizing the hydraulic fluid system.
These overspeed trip systems can be tested while the unit is on-line.
The staff has reviewed these systems and has concluded that the turbine generator overspeed protection system meets the guidelines of NUREG-0800, SRP Section 10.2 and can perform its design safety function.
According to the applicant's inservice inspection and testing program, each compartment of the mechanical and electrical overspeed protection systems will be tested during normal operation, on a weekly basis, by the following tests:
1.
A mechanical overspeed trip test at the EHC Panel to test for operation of the overspeed trip device and mechanical trip
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valve.
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A mechanical trip piston test at the EHC Panel to test for electrical activation of the trip mechanism.
3.
An electrical trip test at the EHC Panel to test for operation of the electrical trip valve.
4.-
A backup overspeed trip test at the EHC Panel to test the 2 out of 3 logic circuits.
In addition, inservice inspection of main steam and reheat valves wil1 include the following:
1.
Dismantle at least one main steam stop valve, one main steam con-trol valve, one reheat intercept valve, at approximately 3-1/3 year intervals during refueling or maintenance shutdowns coin-e ciding with the inservice inspection schedule required by ASME Code Section XI, and conduct a visual and surface examination of valve seats, discs and stems.
If unacceptable flaws or excessive corrosion are found in a valve, all valves of that type will be inspected.
Valve bushings will be inspected and cleaned, and bore diameters will be checked for proper clearance.
2.
Main steam stop and control, reheat stop and intercept valves, and turbine overspeed trip mechanism will be exercised at least once a week by closing each valve or performing the overspeed trip test and observing, by the valve position indicator, that the valves move smoothly to a fully closed position.
This observation will be made in accordance with technical specification requirements by actually watching the valve motion.
General Electric has completed an analysis of turbine missile
~
generation probabilities at destructive overspeed which can serve as a basis for evaluating the adequacy of the applicant's overspeed l
- h. '
, protection system inspection and testing program.
The reports are submitted to the NRC, and are under review by the staff; Until the review is complete, the NRC alternate criteria, described above in this SER will apply to the Vogtle Units 1 & 2.
Design Speed Failure Prevention Failures of turbine disks at or below the design speed, nominally, 120 percent of normal operating speed, are caused by a non-ductile material failure at nominal stresses lower than the yield stress of the material.
Since 1982, the staff has known of the stress corrosion cracking problems in low pressure rotor disks of General Electric turbines.
General Electric has developed and implemented procedures for inservice volumetric inspection of the bore and keyway areas of low pressure turbine disks.
They have prepared and 7-submitted reports for NRC review which describe their methods for determining turbine disk inspection intervals and relating them to missile generation probabilities due to stress corrosion cracking.
General Electric reports are submitted to the NRC and are being reviewed and evaluated by the staff.
Until the review is compl'ete NRC alternate criteria apply to the Vogtle Units 1 & 2.
Summary Tae staff has reviewed the Vogtle Units 1 & 2 facilities with regard to the turbine missile issue and concluded that the probability of unacceptable damage to safety-related structures, systems, and components due to turbine missiles is acceptably low (i.e., less than 10 7 per year) provided that the total turbine missile generation probability is such that the conformance with the criteria presented in O
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Table 1 is maintained throughout the life of the plant, by acceptable inspection and test programs.
In reaching this conclusion, the staff has factored into consideration the unfavorable orientation of the
. turbine generators. The relevant General Electric analyses may be used in determining the inspection interval for turbine disks in the Vogtle Units 1 & 2.
Within three years of startup, no cracks have been observed in a General Electric turbine wheel with depths greater than one-half the critical crack depth calculated for that wheel.
For these reasons, the staff is allowing the applicant up to three years from initiation of power output to propose a revised turbine maintenance program (which establishes, with NRC approved methods, inspection and testing procedures and schedules) and obtain NRC approval of their program.
In response to an NRC request, the applicant has agreed to:
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I 1.
submit for NRC approval, within three years of obtaining an operating license, a turbine system maintenance program based on the manufacturer's calculations of missile generation prob-abilities, and 2.
conduct turbine steam valve maintenance (following initiation of power output) in accordance with NRC recommendations.
Based on our review and this agreement, we conclude that the turbine missile risk for the proposed plant design is acceptable and meets the requirements of General Design Criterion 4.
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1 3.5.1.3.4 References 1.
S.H. Bush, " Probability of Damage to Nuclear Components Due to Turbine Failure," Nuclear Safety, 4, 3, (May-June) 1973, p. 187.
2.
L.'A. Twisdale, W. L. Dunn, and R. A. Frank, " Turbine Missile Risk Methodology and Computer Code," EPRI Seminar on Turbine Missile Effects in Nuclear Power Plants, Palo Alto, California, October 25-26, 1982.
3.
See NUREG-0800, Standard Review Plan Section 3.5.1.3, " Turbine Missiles," Rev. 2, July 1981 for a description of the evaluation procedure previously recommended by the staff.
4.
NUREG/CR-1884, " Observations and Comments on the Turbine Failure at Yankee Atomic Electric Company, Rowe, Massachusetts," March 1981.
5.
Preliminary Notification of Event or Unusual Occurrence -- PN0 -
l III 104 - " Circle in the hub of the eleventh stage wheel in the main turbine" at Monticello Nuclear Power Station, Nov. 24, 1981.
6.
Licensee Event Report No.82-132, Docket No. 50-361 - " failure of turbine stop valve 2UV-2200E to close fully" at San Onofre Nuclear Generating Station, Unit 2, Nov. 19, 1982.
7.
J. J. Burns, Jr., " Reliability of Nuclear Power Plant Steam Turbine Overspeed Control Systems," 1977 ASME " Failure Prevention and Relia-bility Conference," Chicago, Illinois (Sept.) 1977, p. 27.
I 8.
D. Kalderon, " Steam Turbine Failure at Hinkley Point A," Proc.
Instn. Mech. Engrs., 186, 31/72, 1972, p. 341.
9.
W. G. Clark, Jr., B. B. Seth, and D. H. Shaffer, " Procedures for Estimating the Probability of Steam Turbine Disc Rupture from Stress Corrosion Cracking," ASME/IEEE Power Generation Conference
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Oct. 4-8, 1981, St. Louis, Missouri.
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DOCKET Nos. 50-424/425 I
MEMORANDUM FOR: Elfiior A' nsam,' Chief *
[g gBA Licens'ag Branch No. 4 16 Division of Licensing i9 Ig O
FROM:
Victor Benaroya, Chief Chemical Engineering Branch Division of Engineering
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON V0GTLE UNITS 1 & 2 Plant Name:
Vogtle Electric Generating Plant, Units 1 & 2 Suppliers: Westinghouse, Bechtel Licensing Stage:
OL Docket Nos.:
50-424/425 Responsible Branch and Project Manager:
LB #4, M. Miller Reviewers:
S. Kirslis, B. Turovlin Description of Task:
Operating License Review Staus:
Request for Additional Information
- f The Chemical Engineering Branch has reviewed FSAR Sections 5.4.2, 6.1.1, 6.1.2, 6.5.2, 9.1.2, 9.1.3, 9.3.2, 9.3.4, 10.3.5, 10.4.1, 10.4.6, 10.4.8 and Item II.B.3 of NUREG-0737 through Amendment 2.
We need the enclosed additional information to complete our review.
The applicant should respond by July 15, 1984, in order to maintain the current review schedule.
Victor Benaroya, Ch ef Chemical Engineering Branch Division of Engineering
Enclosure:
As stated Conta'ct:
S. Kirslis B. Turovlin x28572 x28556 cc:
R. Vollmer w/o attachments W. Johnston D. Eisenhut C. McCracken M. Miller T. Sullivan S. Pawlicki T. Novak
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S. Kirslis B. Turovlin gOMk
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1 Additional Information Needed from Georgia Power Company for Vogtle Electric Generating Plant Units 1 & 2 The information provided in the FSAR was not sufficient for us to complete our evaluation.
To complete our review, we need the following information:
281.1 Identify all materials (i.e., S. S. type 403, 316, etc.) in the (9.1.2) spent fuel storage pool including the neutron poison material, rack leveling feet and rack frame.
281.2 Provide test or operating data showing that the neutron poison (9.1.2) material will not degrade during the lifetime of the spent fuel storage pool.
281.3 Provide details of your proposed inservice inspection program to
'(9.1.2) ensure that the quality of the racks is maintained.
Demonstrate that the program is adequate to reveal instances of deterioration Q
of the poison materials and metallic materials during the life of V
the spent fuel storage racks.
Provide information on the frequency of inspection and type of samples used in the monitoring program.
281.4 Show that no buildup of gases will occur in the cavities containing (9.1.2) the poison materials.
281.5 Provide a description or drawings showing how poison material is (9.1.2) positioned, held and encapsulated in the rack structure.
281.6 With respect to the secondary water chemistry control program, state (9.1.2) the procedures for recording and management of data, defining corrective actions for various out-of-specification parameters.
The procedures should define the ellowable 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-SR, EPRI/SGOG guidelines).
Because of the significance of condenser in-leakage the chemistry program should include a corrective
--e 4.
, action provision such that a condenser inservice inspection program will be initiated if condenser leakage is of such a niagnitude
)
that power reduction is required (action level 2 of the EPRI/SG0G guidelines) more than once per three month period.
281.7 Identify (a) the authority responsible for interpreting the data (FSAR and initiating action (b) the sequence and timing of administrative 19.3.5) events required to initiate corrective action.
281.8 Regarding the Spent Fuel Pool Cleanup System, describe the sampling (FSAR procedure, analytical instrumentation and the frequency of analysis 9.1.3) to monitor the Spent Fuel Pool water purity and the need for ion exchanger resin and filter replacement.
State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water and for initiating corrective action.
Provide the basis for c..
establishing these limits.
Your response should consider variables j
such as:
gross gamma and iodine activity, demineralizer and/or filter differential pressure, demineralizer decontamination factor,,
pH and crud level.
281.9 Indicate the total weight of electrical cable insulation materials (FSAR that are uncovered and the total weight of cable insulation materials
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6.1) that are in closed metal conduits or closed cable trays inside the containment.
281.10 The information you provided on the Post-Accident Sampling System (FSAR (PASS) is inadequate to demonstrate compliance with NUREG-0737, 9.3.2.3)
Item II.B.3.
Provide information that satisfies the criteria in Attachment 1. provides an acceptable methodology for preparing' plant specific core damage procedures.
Guidance for analytical chemistry procedures is enclosed as Attachment 3.
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t EVALUATION OF GE AND SEC CliEMICAL PROCEDURES FOR POSTACCIDENT ANALYSIS OF REACTOR COOLANT SAMPLES
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Novb ber 1981 1;.
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Idaho National Engineering Laboratory Idaho Falls, Idaho 83401 i
g, The Nuclear Regulatory Comission
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2.0 BACKGROUND
4 2.0 REQUIREMENTS AND EVALUATION CRITERIA FOR THE CHEMICAL ANALYSIS OF REACTOR COOLANT SAMPLEI
............... 6 3.1 Requirements 6
3.2 Evaluation Criteria 7
4.0 EVALUATION OF CHEMICAL PROCEDURES FOR ANALYSIS OF POSTACCIDENT
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4.1 SEC and GE SampTe CoTiection, Recomended Analysis Methodology, and Chemical Procedure Evaluation Program 14 4.2. Chemical Procedure Descriptions, Advantages / Disadvantages and Evaluation Scuaries
................16 4.2.I Baron Analysis Procedures
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7'T; ETectrodei(FSIE) 17 6s 4.2.1.2 Curcumin Spectroph.otometric.......
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4.2.1.1 PTasma Spectroscopy.....
20 4.2.1.4 Baronometry................
21 4.2.1.5 Dig 1 Chem Analyzer of. Manual Mannitol Titrations
..............22 4.2.1.5.. Ion.Chromatogra#iy.(IC) 24 4.2.1.7. C' rminic Acid Spectrophotemetry-....
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4.2.I.8 Conductivity of Baron Solutions...... 27 4.2.1.9 Sumary and Analysis for Baron
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y;y.ggyy;.. q u q.,.
4.2.Z ChToride AnaTysts Methods 32 4.Z.2.1-Ion Chromatography (IC)
.........32 l
4.Z.2.Z: Specific Ion Electrode (SIE}..
34 4-MT Turt:1dimatric,, Colorimetric,. Titrimetric,
~ and Spectrophotomatric
......35 1
&.Z.2.4 Conductivity of Chloride Solutions 36.
4.I.Z.E Sumary and Conclusions of Chloride Analysts Procedures.
37 4.Z.7 Dissolveci Hydrogen and-Oxygen 37 4.2.3.1 Gas Chromatography (GC) - Hydrogen Analysis
...........;...37 4.2.7.2 Gas Chromatography. Yellow Springs Analyzer -
- Dissolved Oxygen Analysis 43 4.2.3.1 Evaluation Sumary of Dissolved Oxygen -
3 and Hydrogen. Analysis
.......45
/
if
~
- - = -.
- 4 :.
w.
t 4.2.4 Conductivity and pH
.................. 46 4.2.4.1 Conductivity 46
~
. 4.2.4.2.pH 48 4.2.4.3 Sumary of Conductivity and pH Analysis 1
Methods 50
5.0 REFERENCES
.............................51.
TABt.E5 1.
Stations with Proposed usage of SEC and E P,ostaccident rampling System 5
~
2.
Sumary of Requirements for Postaccident Chemical Analysis of Reactor Coolant 3amples,
....................... 8 3.
SEC and E Reactor Coolant Analysis Methodology 15
~ -
4.
Chemicai Analysis Procedures Cons'idered by '5EC and E 17 S.
Features of Proposed Analytical Procedures for 5cren 29 5.
Features of Proposed Analytical Precedures for Chloride
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....,-.,..a EVALUATION'0F SEC AND GE;ANA1.ITICAL CNE$ICAL MOCEDURES FOR POSTACCIDENT ANALYSIS'0F REACTOR. COOLANT SAMPLES LO. '.
SUMMARY
A!!D' CONCLUSIONS Summary
~ As a rtisuit of the Thres M1Te'IsTand Unit 2 incident, ~ the Nuclear Regulatory Comission (NRC) required licensees of nuclear power plants to irplett, by. January I.,1982, the capability to collect and analyze
~
~
reactar coolant sareples foTTowing an accident.
A nuirber o'f licensees have proposed the use of.postaccident. sam' ling and analys_is. systems sup'-
~
p plied by Sentry Equipnent Corporation (SEC) or General Electric Comoany (GE).
Under a. technical assistance contract to the NRC,, Exxon Nuclear Idaho Compary, Inc. (ENICO) evaluated the. sample ecliecticn and chemical analysis procedures associated with the.two systems.
The objective of U
the evaluaidon 'was to determine Ipplicable procedures and to identify F*
the most appropriate method...The study.. involved. a ' review of - the NRC
.... ~..z. ; w.
..c....
requirements, the estab'Tishment of review criteria,, and the evaluation of the proposed anaTysis methods and test. data against the recuirements and evaTuation criteria.
~
The~ most appropriata methods selected by? ENICO for the requiied chemicaT analysis of postaccident reactor coolant samples are shown be'-
icw.
Detailed descriptions., advantages, disadvant ses, and/or defi-ciencies.of the selected procedures aie sumarized.in section 4.L Also in section 4.2 is the same information for other procedures proposed by SEC'and-GE It is worthy of note that a nunter of the other procedures proposed are also appropriate,. as indicated; included below are only
~
those deemed most appropriate.
L. Baron ' Fluoro'!! crate Selective Ion ETectrode 2.
Chloride - Ion Chromatography
' [
I.
Dissolved Hydrogen - Gas Chromatography 4.
Dissolved Oxygen - Oxygen Probe 5.
Conductivity. - Conductivity Cali 6.
pH - pH Probe L
1 l
=
=
)
A1though ENICO *did not conduct tests to evaluate the suitability of any of the, procedures; in ENICO's judgemeit, the laboratory tests per, formed by SEC and E are sufficient to provide t. high degree of anurance of 'the suitability of the ie'1ected and tb. noted alternate ' procedures for analysis of accident reactor coolant samples.
For suitability testing of additiona1 analytical procedures, ENICO recommends that standard test matrix samp1es be utilized to demonstrate theit" acceptability. _ Standard matrix solutions similar to test solutions
~*'
'empToyed 'by SEC are recommended as ' they contain the most s'fgnificant core ' degradation. products in ' oncentrations equa1 to or greater than c
'those ' projected"froi:r an a'ccfderit with a liegdla~ tori Guide i.3' or 1.4'
~
source term. Test solutions used by SEC consider the effects of chemi-cils which might be added to the reactor cociant following an accident.
For chemicai proc $tdures that are to be used for the analysis of undiluted reactor cooTant sampTe's,. the following standard test matrix C
containing nonradioactive species is recc. mended.
Constit'uient Concentration' ( ocm) t-
. 40
.._.....3....
A Ba 10 t.a*3 5
~
- . m. e 5a
- ,.gc,,
.s C1-
~
10 B'
2000 Li 2.
N0f
~
150 NH+
5
~
K N
~
S
.,e
\\
)
7 - _,
_ e _.. _
.o.
?
For chemicai procedures.that are to ha used for analysis of diluted reactor toolant samples, ' testing ishould -be. ' performed with a stand $rd
- s.,
satrixldf'Tuted by, e volume. aqual'to the dilution' to be used in the proca-dure to be tes.ted.
It is also recommended that the procedures and' as-sociated instrumentation b's! tested in 'an induced ganna radiation field which wf 7T yield a. total absorbed dose of 104~ rads per gram of reactor s
cociant.
7 i
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2.0.8ACKGROUND
..From studies of t.he incident at Three~ Mile Isl,and Unit 2 (TMI-2),
1
~
the need for 1mpr'avement'of' the capability'of licensees of nuclear' power plants to determin$it p1' ant conditions in a more timely' manner was identi-fied.
Subsequently, the NRC issued.1-5 for implementation by the
~
' licensees, specific requirements in several areas for improvement of the capability.
In addition to 'the developnent and implementation of the upgraded capabilities, the requiremenf.s specified that the licensees shouTd prepare and have naiTahTe documentiation of the capabilitie~s for a post,-implemen.tation evaluation,of compliance.
~
Exxon Nuclear Idaho Company, Inc. (ENICO) was can'tracted by the NRC's Division of Licer. sing to provida techn'ical assistance for the eval-uation of the post-iaplementation documentation in a number of areas.
One area was *Postaccident Sampling Capability", Item II:3.3 of NUREG-0737.6 It pertains to the ab1Tity of the licensees,to cbtain reacter c
r co61 ant and containment atmosphere samples and to a,nalyze the samcles D-for selected radionuclides and chemical jspecies under accident ecnditions.
.[
In order to facfTitate tfie evaTu'ation of the post-implementation I
documentation, ENICO was requested ta evaluate the applicabiTity of the chemical. and radiological anaTpis capabilities associated with two.
8 postaccident 'sampffng b2aproposed,: for use by several power i Tantr'
~
(TahTe 1)
The two system vendors are Sentry Equipment Corporation (SEC) and General ETectric Company- (GE) 9 The initiar plan called. for ENICO to evaluate the-SEC system only and to perfornr the evaluation in twar phases As it was believed that current technology was suitable for radiological analysis, the two-pnases were to ha a brief sunnary report on the chemical analysis proca-dures and a more detailed report on both the chemical and radiological analysis procedures.
However, due to manpower shortage at ENICO and an NRC request ta incorporata the GE system into the evaluation, an alter-
)
i nats approach was taken.
The alternats approach is to: 1) evaluate and prepare a detailed report of thr chemical' procedures for both the SEC 4
j
~.-
~ _
.u~2;
.-_.g.
.t TABLE 1
' STATIONS WITH. PROPOSED ' USAGE.0F.5EC AND GE POSTACCIDENT SAMPLING.SY. STEM ENERAL ELECTRIC SENTRY Brunswick 1/2.
Dresden 1/2 Itine MtTe Point 1
. Quad Cities 1/2 Fit:: patrick Zion 1/2 Oystar Creek Browns Ferry 1/2/3 PtTgrim I SaTem 1 Duane Arnold Kewaunee
.tnticalTo Indian Point 2 Peach Bottc:s 2/3 Surry 1/2 a.
North Anna 1/2 Palisades and GE system and 2) avaluate' and document the radioche:nical analysis
-t) procedures associated wit'h both sampling systems later.
The detailed
, - eva-luatiien 'of the' chenciat analfifs' procedures'.'i?s. the topic M..this'..~
'i report, which is Tirsitad to the anaTysis of reactor coolant ' amples.
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3.0 REQUIRENENTS^AND EVALUATION CRITERIA'FOR THE CHEMICALANALYSISOFREACTORC00LANT'SAiPLES
~
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~
3.1: Requirem..ents To provide information for the assessment of core integrity, sh'ut-down neutron adsorber concentration, and reactor coolant corrosion poten-tial; Ticensees or applicants for licenses of nuclear power plants are
~
~
required to establish a capability for the timely collection and chemical' analysis of reactor. coolant samples under accident conditions.
Per
" NUREE-0737 th'a required i:hemicaT analyses for ' reactor coolant samples
are baron (PWR only), chloride,. arid either t'otal dissolved gases or hy-
'erogenithEmeasunmentofdissoTveif oxygen is recommended in' NllREG-0737' 10 but not required.
Per Regulatory Guide 1.97 the measurement of dis-solved oxygen, pH, and baron in all plants *is required.
NUREG-0737 alse
~
specified that the analysis could be performed by employing a combination I
of pressurized /unpressurized, diluted / undiluted grab samples or inline monitoring methods.
However, for analyses performed by inline methods,
('I s.
a capability to collect backup grab samples and' to provide procedures O
for their analysis is required.
In all cases, the collection of grab
' samples'for analysis and the"inline ana' lysis must be ~able to be perfoimed with or without the operation of arr auxiliary resc' or cool' ant system, t
~
e.g.
Tetdown.
('
' With the exception of the. chloride analysis, the time allotted for sampiing and on'. site anaTysis' of the~sampies is.three (3) hours or Tess.
~
Time aliotted for the chloride analysis, which can be performed offsite, depends on the type of reactor coolant water and the nunter of barriers betwee'r the reactor coolant wa'ter and the primary containment system.'
r For plants with seawater or-brackish reertor coolant water or with a single-barrier, primary coolant containment system, chloride analyses are required withirt twenty-four (24) hours.
For other plants the chlo-
~
~
ride analysis is required within ninety-six (96) hours.
In addition, the licensees or appitcants are required to consider the radiological hazards associated with the sample collection and analy-
)
11 2
ses.
The assumptions of a Regulatory Gide 1.3 or 1.4 source I3 term and radiation exposure 11mits of fiva (5) rem to the whole body 6
..,.c.._......
or seventy-five (75) rem.to the extremities of any individual are'to be used in system" design and selection of chemicaT analysis methods.
Last,'the licensees.or applicants are to provide provisions fo'r
~
restricting background radfation TeveTr irr the. chemical analysis' facility and for insuring the validity and accuracy of the'saniple analyses..These provisions; include such things as sample shielding, adequate ventilation air and filtration, proper sample disposai, sample line purging,' reduc-tion of plata-cut in sample lines, etc.
.wy
..- -?. _ ;
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The requirements for post accident chemical analysis of reactor coolant samples are presented. in Table 2..
3.2 Evaluation Criteria The chfew-tive of the present evaluation of potential methods for the chemical analysis of reacter coolant samples under pose 1ated ac$:1-dent conditions is ta determine-4p'plicable procedures and to identify the most appropriate procedure.for each of the required analyses.
n Q
i
.Many factors were:. considered.
n t.h.e eval.ua. i'a.n of,the pro,po.s. ed meth-t.
ods. Obviously, compliance with the requirements of sansitivity, accura-cy, range,. analysis' time, radioTogicai ' dose.Timitaticas, and sampTe cot-Tection methods were evaluated.
This included comparisions of the advantages (Tower. radiological exposures) and disadvantages (reduced,
sensitivity 'and' accuracyf af' ut1Tizirig "dfTuted or very small
~
reactor coolant samples versus Targar undiluted reactor coolant samples.
It also involved an estimatierr of~ the significance of chemical and radio-logically-induced interferencer.
Othei-factors which were considered are the c'omplexity of' the procedures and the applicahility of the tech-nique te both accident and normar conditiert usage.
.Due to tha unavailahility of informatiert in a nunter of ins *mcas, factors not considered were specific design featurcs of the two sampling systems. Examples are sampling. locathns, shielding, sample line purg-ing, sampia validity, ventilation, etc..
7
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- TABLE 2
+-
SUMMARY
.0F. REQUIREMENTS FOR POSTACCIDENT CHEMICAL. ANALYSIS-
~
~
~
.0F REACTOR COOLAN.T. SAMPLES i
01ssolved Gases (1)
Analysis Capability Baron Chloride Total -
Hydrogen Units (ppa)
(ppm) sec/kg sec/kg 1
Raouirement' 0-1000((3)
G-20 0-2000~
0-2000
. Range 0-6000 4)
~
Accuracy.(6)
Percent 4 if >1000
+10 1f >0.5-
+10 if >50 +10"if >50-Units B G if <1000 70.05 if <0.5
~+5 ff <50 9_ if <50 Sampling)
MethodD Inline Optional Optional Optional Optional Grah Sample ~
Regiured Required Required.
Required
- Analysis
!Q Location
~
Onsite Required Optional (8)
Required
- Required Offsite Optional Optional Optional Optional Samoie' 24(8)
~
Colleucion I
3 3
and Analysis-
- -p w.
t- '--- 96 Time.(hours)..
RadioTogical..,n g.,(91 I;.
(Sh.
.(9). -
(9).
2 3
Exposure
~
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Notas:
I.) A pressurized reactor coolant sample is not required if the dissolved: gases can be. determined with an unpressurized sample.
- 2) The measurement of conductivity is. not required in KUREG,-0737 or Reg. Guide 1.97, Revision 2; however, methods to measure conductivity are proposed by SEC and GE.
Accordingly, the measurement of conductivity has been included in this study.
8
=
=
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e.
TABLE 2 (Continued)
SUMMARY
OF REQUIREMENTS FOR. POSTACCIDENT CHEiICAL ANALYSIS OF REACTOR COOLANT SAMPLES Analysis Capabiiity Ox' gen pH Conductivity (2) y Units (ppa)
(pH units) pS/c:
- s Recuirement
~
Range 0-20
' I.-13 1-1000(5)
~ ~
~
Accuracy (6) -
Percent
+10 if >0.5 Not appl,1 cable
+ 20
. Units
+0.05 'if '<0.5.
+ 0.3 ff 5 >pH<9 Not Applicable
~ '
'~7 0.'5 if~.5'<pH)9
~~
Sampling)
MethodU Inline
.... Optional Opti6nal Opticnal Grab Sample' Required Required Rees4 ped Analysis Location
. ;),..
~-
Onsite Required Required
.Raqu4*e4 Offsite
' OptionaT OptionaT Optional
. :s Sample.
. 3 --
ca3zeeg3ba u
.7,....:..
3
. c.
and Analysis Time (hours) c Radiological (9)
(9)
(9)-
Exposure Limits.
- n Notas:
(Continued) 3)-Boiling Water Reactors
- 4) Pressurized Water Reactors
- 5) The required range for measurement of conductivity was taken from reference 14.
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'6)'The' designatfori' of ' percent ~ ~ accuracy as l' 5 if ~>1000
... indi. cates th.e required accuracy. is','+
!I per,c.an.t.. if.. the. *
... :t. s
, required measurement,is greater than 1000 units.
The
. designation of units accuracy as + 50 if' <10' 0.. indicates 0,'
that the requfred accuracy is + 50 measurement units if the required meast,rement is less than 1000 units.
The required accuracies were taken froir reference 7.
7)' Analysis may be performed with either grab sampling or inline
~
~
monitoring esthods.-
However, 'for -infine ' arralysiir methods.
the-capability to coTTect' and ' analyze backup yab samples is'
~
-' ~
.. ~ -
.. ~.
requir..ed.
The capability to collect and analyze at least' one sample ter day for seven (7) days following the onset of the accident. and at least one sarispie per week until the.ac-cident no. longer exists is also required.
. 8) For nuclear power plaritis which utilize seawater ci brackish '
['
water as a ' source.'of' reactor coolant ' water or which have a. *
, single-barrier, reactor ecolant containment system,.the chloride analysis must be performed.within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />si fer other nuclear power plants, the required chloride analysis--
time is 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
T.te chloride anaTysis may be performed-offsite.
, 97 The radiatforr' exposures tcr any individuai' involved in the co1Tection and analysis of reactor coolant samples under
/
accident conditions may not exceed. 'S rem to the whole body or 75 rest to the extremities.
e i
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The evaluation critieria used in the study were:
. ~.
1.
AnaTysis. Time.
.As, the time required for.. sample. co.llection was 3
y not specified by SEC.or. GE it was assumed.. sample collection could be performed within one hour.. Accordingly, an upper limit of two hours was allotted for sample analysis; chemical analysis wares whicfr required 'two hours m-Tess for analysis were
)
e satisfactory.
~
=
. Z..fansitivity, Range, Accuracy. - Chemical procedures which encom-passed the entire measurement rang'e with the required accuracy.
were considered adequatal ~ To cover the fuli range of measure.
ment as required, sample dilu* tion methods were considered
. satisfactory.
r -
.,5
. e : :: : '.
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..... a i 3.
Radiological Exposure Limits, Sample Size. - Radiological.exco-sure to any individual is limited to 5 rem to the whole body and 75 rent to the extremities during.the collection and analysis n
.J of reactor coolant samplet:.
Under the assumption of Regulatory v~
_~.,.~; Eu; ides. f T or,.1;.'4,.
leas,es of -fission.. products.to/the. reactor..
cociant, cal.cuTated dose rates from reactor. coolant samples are nominaTTy I40 R/h/g at'10 cm with a one hour decay.
Main-tenanca 9f radioTogical exposures within acceptable Timits re-quires the usage gf safety fa,c.. tors suctr. as: shielding, distance, I
3.;..-
3._..
.s.
-e..
. s.. -
exposure times sample diTutton,. very. c. r.small undiluted' sample, 1
l and/or inline monitoring.
l I
.The chemical analysis. procedures, including dissolved gases, proposed by GE and SEC make usa of irtline monitoring very small undiluted samples,. and remote diTutions of the initial reactor c'colant. sample.
The diluted. reactor coolant samples are used for subsequent'
- hands-on" anaTysis.
With the exception of the subsequent hand!-on analysis, this study did not evaluate the radiological hazards associated with the above methods.
It was
~.
assumed that adequate shielding and/or remote operation would minimize radiological exposures to personnel.' In regard to the subsequent analysis of diTuted reactor coolant samples; only estimates of radiological exposures could be made as they are 11
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not only.a function of the amount.of reactor coolant in :the-sample, but also depend on the techniques of the analyst and
- - ' the desigii of' Oie arialytiiAl 'fac1Ti't9
~
The method which w'as' established to lessen exposure is' to limit the amount'of reactor coolant in the sample taken for analysis to 0.1 mi.
The basis for this criterion,is the knowledge that doses to the extremities will be the limiting factor for hands
....~..,
. example, calculated exposures to on chemical. procedures.
For.
the extremities,, using the above value of 140 R/h/g of reactor'
. coolant at.10 cm, ' wit. T. exceed t. he 76. rem limit. by' a. factor of 1
almost two for a two hour exposure to a 1. mi sample.
It would require approximately ffve hours of continuous exposure to ex-
~
ceed the dose limits for a 0.1 mI' sample.
It-is realized that the limitation to a 0.1 mi reactor coolant sample size is con-servative as exposure t[ime will, in reality, be less th'an two
,bours and techniques to reduce the exposures will probably. be employed.
However, to allow.a sufficient margin 'of safety, a 0.1 m1 reactor coolant. sample was considered an acceptable size.
' sample in this study.
In a final evaluation of acceptable sam-
' pie sizes [ Targer samples may be permissible,.but aiT. factors must be considered.
c..
- 4; compTeittyZ. Routine /Accidditt Us' age
.Thi:E'ather"cNtaria which
~
were used are the complexity of performance ~ of t'he proceduies and the applicability of.the procedures to both routine and
. accident conditforr usage.
The procedures were assigned low, mediunr,. or high levels of complexity based primarily. on the number and naturn of' manipulations involved in the procedure.
Procedures with applicability to both routine and accident con-
~
ditions were considered more satisfactory than procedures ap-p11 cable to accident conditions as the use of nonroutine proce-dures can creata confusion and cause errors under accident conditions.
)
5.
Chemical and Radiologically-Induced Interferences - The release of large quanities of both r;dioactive and nonradioactive fis-sian products will result in high radiation f'ields sind chemical-
w......
o.
- v..
ly significant levals of various ionic species in tha reactor coolant.- Both the radiation. and ionic species can einterfere
- with the accuracy of chemical pr%cedures ~used 'te analyza reactor cooiant samples.
In the. selection of 'an ' appropriate chemical.
analysis method, 'these~ mitrix ' effects should be considered.' 'In thiir study a chemicai procedure w'as considered unsatidactory if the interferences cause the. accuracy of the precedure to
' exceed he required Ti$iits. " The ' evaluation tricluded a review t
of available test data and professional judgements based on past experiences of personnel involved in the' review.
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.. 4.0. EVALUATION.0F QiEMICAL P.ROCUDURES.. EOR ANALYSIS, OF POSTACCIDENT REACTOR. COOLANT SAMPLES.
/ Inithe evaluation of the.'applicah1Tity "of chemical ' procedures 'for
~
anal sis of'postacefdent reactor coolant samples, ENICO studied the ' chem-istry of tha procedures, compared their capabilities with NUREG-0737 requirements and. the established evaluation criteria, and ranked the
~
~
procedures in order of appropriateness. Some of the procedures are simi-1ar to ones used at the Idaho National Engineering Laboratory (INEL);
this experience a'dded i:o-the data base.
~
i resented below 'irr Sedtic'ri '4.Il 'ts a ' summary' of"the samp1~e collection
~
P
~
(
and chemical analysis procedures proposed by SEC and GE and a general outline of the testing program conducted by' SEC and GE.
This is followed
'by a. presentation of ENICO's evaluation of the procedures.
Included are brief descriptions of the procedure methodology and_the aByantages and/or im disadvantages of each procedure 7 ; Last, 'tha overall evaluation of the
{
, individuai precadures are summarized for a given ' type analysir..
4.1 SEC and GE Sam' ole Collection, Rec::rmtended Anaiysis Methodolocy and" Otemical Procedure Evaluation procram Methods for analysis of postac=ident reactor coolant samples pro-U M-18
~
posed by SEC and GE include irritne monitoring and laboratory angTysis of ' grab sampTes.- For Intine' monitoring,.,sampia stieams are diverted e'ither continuousTy or intermittently through intine sensors.
For laboratory analysis,. the reactor coolant grab samples are diluted inline before transfer to the, Taboratory or directly to an analytical instrument. Either d.iluted Tiquid or dissolved gas grab samples can be obtained Ta obtiairr Tiquid samples,.1.:100 dilutions of 0.1 m1' reactor j
- coolant samples are. typically performed;. larger initial dilutions or secondary dilutions of the initial diTution can also be performed.
To l
obtain a dissoTved gas grab sample; thirty (30) to seventy (70) milli-liters of pressurized f sactor coolant are isolated, the sample is depres-
^
surized, and the dissolved gases are purged into a gas holding chamber
)
with an inert gas.
One millilitar or larger aliquots of the diluted sample are analyzed following filution with the inert gas to a knowr.
pressure.
. L - - _.
- - = - = -
=
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=
=
v.
~
.The chemical, procedures associated with. the. proposed methods ~ are either conventional or modifications.of conventional, chemical. analysis t ~
procedures.- A summary of the methodology; inclucing chemical analy.is
.;,cer.edures. reconnended by SEC and suggested by E is presented in. Table 3.
Not included in Table 3 are other meth'ods de. tailed by SEC.and GE; this information is, included in section 4.E.
..., -. -,.. :..r.
m TABLE 3 SEC AND GE REACTOR COOLANT ANALYSIS lETH000 LOGY
.=
SYSTEM.
SAW LE ANALYSIS ANALYSIS VENDOR..,
. TYPE METHOD Baron SEC Grab FTuoroborate GE
. Electrode Erab Spectrophotometr'ic (carmlinic acid)
SEC
.Infine, grab
. Ion Chromotography C,}
E Turbidimetric V'.,.
s
, ~,
c..
n.
..,p....
Dissalved SEC
....... _Inline, grab Gas Chromotograpny
. :. a.
.e Hydrogen E
~ Erah Gas Chromotogra#1y'
[.
Dissolved SEC Inlina YSI 0xygen Analyzer.,
0xygen E,
Grah Gas Chromotography 2_
~
Inline pH probe pH SEC E
Grab pH paper l
i Conductivity SEC Inlina Conductivit'y Cell E
Inline Conductivity Cell i
n O.
t
(
s 15 i
T w C y *:.
3.}
~
......w l
,. Chemical procedures recomended by.SEC 3re..the.. result of a.. develop.,
.l ment and testing. progrant. conducted by Nuclear Utility Services..(NUS) for '
SEC'. r "In the. study' the recommended methods and cseveral other chemical. -
- -r canalysis' methods were evaluated in the 'Taboratory to' identify chemical interferences due to sample matrices, to determine operational charac--
teristics of instrumentation, and. to measure the sensitivities, ranges, and accuracies of methods.
Employed in the study were simulated post-accident reactor coolant test samples.
They contained, in addition to the chemical species of ' interest., high-yiaTd, stable fission products and'~approprfste conceiii: rations of chemicaT additives anticipated te be
~
~
present in the reactor coolant foTTowing an accident.
The study did not iriciude actuai' measurements of possible effects of nigh radiation fieTds-on the procedures;. however, it did include the results of a survey of I
personnel with prior experience' in the. analysis of samples with high radiation fields and a literature review of effects of high irradiation on different materials.
.r, '
(v) '
ChemicaT procedures suggested' by GE, except those coincidential to
~
the.SEC tested method.s, are not the result of detailed laboratory test- '.
ing.
The only testing of the procedures is related to the effects of high feradiation.of.the sampier.. The suggested procedures were selected primarily' on the basis of simplicity, stability and availability of re-agents, minimunt radiation exposure; and likelihood of causing contamina-tion problems.g.
7.
... - 4 :f r;..-
4.2 ChemicaT Procedure Descriotions, Advantages / Disadvantages and Evaluation Sunmaries In 'the stTection,. rem-sndation, and/or suggestion of chemical procadurer for analysis, of postacciden,t re. actor coolant,. SEC and GE con-sidered. a total' of twi.j-seven (27) procedures. The chemical procedures considered by the twa vendon are shown in Table 4; also noted in tho s
table are known procedures in use at the INEI.. As many of the procedurcs
~
are similar or identical, they have been grouped together, as appropri-ate, in ENICO's evaluation of the, procedures.
presented in order below are the evaluations of the baron, chloride, hydrogen, oxygen pH and con-ductivity measurement procedures.
e
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,....:. c.
- v.n....,
E" y,.
(
' h.. ? - Om. QfEMICAlsANALYSIS' PROCEDURES CDNSIDERED BY SEC AND GE f
CHEMICAL ANALYSIS-VENDOR PROCEDURE' 4
Fluorocorate sele ~ctive ion electrode
.' 5EC Baron
~SEC Curcumin Spectrophotametric
- E..'.., 1.Baronmetry*R1asma Spectroscopy.,..
.l.
- e.=O2 d4 P~
SEC SEC
'. Digi Chem Analyzer MannitoT Titration
~
SEC, GE Cartninic acid Spectrophotometric SEC Ion chromotography*
- ,:.,. n:. p-+ mgyg.%'i:n 'Nanual Mannitol Titration
- E Conductivity of Baron Solutions ChToride' l!
~SEf F-
' Iorf Chromatography * '
SEC Selective Ion Electrode.
SEC Mecuric Nitrate Titration SEC, E Thiocyanate Spect!rophotometric
'... ' E StTver Chloride Colormetric E
Conductivity of Chloride Solutions Hydrogen,
GE,SEC Gas ' Chromatography
~
~
E Eas Chromatography Oxygen.
- i,g.
SEC YSI 0xygen Probe s,
GE't::: ud:n pH. Paper :
- p H... ~,
- v. x s..
GE o -...z.,. Conductivity ggg
...r.
-' $ Probe *'
~
Conductivity 9 SEC,'GE :
' Conductivity Call *~
' %.
- 1_..-].,
w Wo%f i '.i h: *L:: :.. -
~
Indicates a procedQre that is in use at the Idaho Chemical Processing Plant (ICPP),or the Loss, of Fluid Test Facility (LOFT) at the INEL.
4.2.1 Bororr AnaTysis Procedures-c.
4.7.1' I* FTuoroborate S'eTectivei Ton Electrode (FSTE).
In the FSIE chemicar anaTysis. procedurei, the baron content of a sampie is
- i
' determined. by the measurement.of the concentration of the tetrafTuorobo-rate fon.
In' addition to the~ sensi'ng eTectrode, which contains a mem-brane with a selective tetrafluoroborate ion exchanger, a single junction reference electrode (KCe/ saturated AgCe) and a conventional millivolt
.- meter 'with a relative miTTivoit made arv required.,
(
c 17 h
1 l.,..
2.-
u.;
- _
a
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.y
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.The, pcedure requires precise,,latyaratory. techni-2.
- ques; care must be. exercised to add, the reagents to.the standards or
. samples in sequence and.to perform-measurements at. prescribed times. <In the analysis. procedure a standard and.a. sample are ' analyzed simultaneous Ty.
Initially,1.0.m1 of' saturated sodium fiuoride is added to 5.0 ml-of the standard, and,then 0.5 ml 10 N_ sulfuric acid is added (the sodium fTuoride and sulfuric acid converts horic acid to the tetrafluordbarate ion). With the addition of the. acid to the standard, a timer is started;
~
five minutes later the same reagents are added to 5.0 ml of a previously
.diTuted sampie... -At eight. minutes.-the electrodes are ' inserted into the standard solution which is being stirred; at ten minutes the millivolt response is" adjusted 'to correspond " to a' s'pecific' value 'on
~
a pre-established calibration curve.
I The miTTivolt response for the sample is recorded at fifteen minutes and related to, the ppn baron from the calibration
,e.
curve.
\\
v j
's.
.3 G.
To minimize radiological hazards LO mi. samples and
'. standards can b'e analyzed by the above procedure with the use of coi-respondingly less sodium fTuoride.af. sulfuric acid.
In addition, the d
analysis can be performed by using onTy 0.3. mi of the originai 5.0 ml
~
or I.0 mi of sample taken for analysis.
The analysis using. 0.3 m1 is performed. staticaTTy frt aferodishes.
There are two typer of ca.libration curves.
One for the 5.0. al. and/or 1.0-mi samples analyzed. by finnersion of the electrodes inta a. stirring s,olution,, and one for the G.J nr1 sampler analyzed by im-mersion of the electroder irr the microd.ishes.
The calibratierr curves are estah11shed using the same techniques. employed for the samples;. the
~
calibration curves are valid only for the pair of electrodec used to estabTish them.- Calibration curves are estimated to be valid for six i
month'; however, frequent use of the electrodes shortens their life.
s Accordingly, routine' checks of_the calibration curves are recommended to
)
maintain their currentness.
Apprd'xima.tely a total of one hour is re-quired, to generate new calibration curves for both large and small samples.
O n
__,=
_ = _ -
3.., e..
, Numerous Taboratory, tests were carried out. with-simulated postaccident matrix samples.to. identify chemical interferences to the-FSIE procedure.
No sample matrix effects were observed when the
. procedure described above was followed..
The advantages of the procedure are its wide maa-surement range and accuracy?the-small sample sizes required,the Tack of '
chemical interferences', its adaptability to routine and accident condi-tion usage, and the short analysis time required.
The main disadvantage of.the procedure is its rela-2 tive~ complexity, which 'w11T' necessitate 'weTT-trained analyst's and fre '
quant usage of the procedure by.the ancly?ts in arr er to retain their I
d familiarity with it.
Another limitation of the prc cadure, under the assumption of a minimum 'inittai sampie diTution of 1:100, 4 the inabi1}-
ty of the procedure to measure baron levels in highly radioaccive reactor
-~
coolant below fifty ppm.
HowevIr, in ENICO's opinion, this is noe a serious ~ Timitation ' as under accident conditions the concantration 'of
_.)
, boron in the reactor coolant should b.e much higher than fifty ppm; and, eif~ it1tsn't;, confirmaf. ion,i:tiat boron. levels areififty ipir..or.above. is.
e sufficient informatiers tz determine.the need for subsequent corrective
~
actions p' '
.., The FSIE anaTysis l procedure has nc~t been. ' ' ested-t witfr high radittien field samples; however, ENICO does not believe ir '
radiation associated with highTy radioactive samples will significantly alter the applicability of the method.
(.Z.1.7 Curcumin Saectrochotometric.
The curcumirt spectro-photmatric baron analysis method fr based on the measurement of a red-colored product, rosocyanine, formed. by the reaction of boren and curcumin To perform the measurement the '1.G al diluted sample and standards, which are analyzed concurrently with the samples, are mixed with'4.0 ml of curcumin;. evaporated to dryness; dissolved in 95 percent i
isopropyl alcohoT to a total volume of 25 al; and transferred to a 1.0
(
cm spectrophotometer cell.
In the spectrophotometer, a Bausch and Lorrb Spectronic 20 or equivalent, the percent transmittance of the sample and 19
. ~
........-l**~~~
,,.. u. --..... a.,
standard kare. measured at 540 nm.A calibratio.n chart is. prepared from i.
-the stan.s t
dards', and th'e concentration of.' he. boron in.the sampia is deter.-
'ained'..from the.. cal.ibration chart..
~
.s The curcunin spectrophotometric procedure was labo-.
ratory tested by SEC/NUS.
In addition to sample matrix effect studies us,ing. samples containing selected nonradioactive, fission products and chemicals-anticipated to be present after an accident, experiments were performed to optimize the ' precision, accuracy, and required analysis
~
j time F
- ?
~
-+-
c a
,.- n
.The advantages of the ' procedure are 'its wide. mea -
~
surement range, its accuracy, the small sample size required, the lack of chemical interferences, its utility under,' accident and routine condt-tions, and its relativ'e simplicity.
The disadvantages of 'the procedure are the long analysis times required, and the necessity to. generate. cali-
.... bration curves at the same time sample analyses are performed. The later f-is'eqnsidered a. disadvantage as a significant amount of. time could be h
wasted if a satisfactory calibration can not be obtained the first time.
Another limitation of'.the Nocedure, is the'.inabi.lij:y to' measure ' ieveis of. boron Tevels in reactor water below twentv com.
However, as noted
,4- ---
i above, SICO does not consider this a major Timitation as required cor-
- 3. p rective. action can be made based on the knowTedge of baron concentrations 7,,., _
oir twenty p;st or, more.:
.y.,.p. y.. g. r a m.
- ~
.g ;
a
. a-The effects of high radiation fieTdr on tha' proca-dure have not been determined.
In DICD's. judgement,, the accuracy or sensitivity of the procedure wou.1d not be compromised; but this needs to be. confirmed. before the procedure is used 4.2.I.3. Plasna spectroscooy The analysis. of boren by plasma spectroscopy ir achieved by vaporization of the sample in a plasma-jet and analysis of the atomic emiss11n spectra which is generated.
The baron resonance wavelength of.either 243.7 or 249.8 nm is used.
Readout
\\
of the unknower is compared to standards.
Five milliters of a diluted reactor coolant sample is required One milliliter of the sample and i
associated radioactivity is completely vaporized and released; the other m
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four milliliters are col,lected in a waste ' container.as condensed spray droplets. 'The required analysis time is fifteen to thirty min,utes.
~,
No specific laboratory testing. details were J pro-videdi however, it was indicated that Timited tests were performed on simulated reactor matrix. scTutions with satisfactory reproducibility and accuracy. The measurement range associated with the procedure also was y
not provided; however, a lower detection limit of less than 1 ppn boron,
_ Z is reported.
With this sensitivity and appropriate sample dilution, it
~
_7 appears the measurement range would b'e sufficient to cover the measured ment range required.
-m. -
~
The advantages of the precedure are its apparent
' simplicity, time required for analysis,.and smal-1 sample sizes.
The disadvantages of the procedure are the lact of
~
sufficient laboratory testing en('the radioactivity releases associated I with it.-
It is assumed that appropriate design mcdifications could be e,
J incorporated to circumvent this latter deficiency;. but the design must I
incTude features.to collect 'all the. radioactin releasas, :not merely..to' u.,.
~
contain' them in a fumshood,' 7as is done with the existing design.
High
~
~
radiation fields EfTT not affect the 'appTicabiTity of the procedure.
s
~ ' ~ ' '-.. 4;Z;I.4 Baronometry; "Tha'anaTjrsis of baron by baronometry is based on the attenuation ;of~ a' ccTTimated neutron beam by a seTution of baron betweert the sourca of neutrons and the detector.
The neutron count rate.from the detector tube it converted directly to boren concen-tration orr the readout electrometer-or pulse counter.
Californium-252 -
or plutonium-bary111ust are typically used. as sourcer of neutrons.-
Baron trifluoride (8F ) tubes or fission chanters are two types of detectors.
3 Although 8F tubes have been reported to operate satisfactoriTy in 3
gamma-ray fields up to 100 R/hr later baronmetert use fission chanters as they are virtually insensitive to gama-ray fields.
~-
m 2I
-l-a??N:W.??
- 4 *
-N-=..
~
~ ~
Boronometers typically employ.
relatively large
')
' vol$ sis s'amples['l-Z '11'teid br more."' 'Ai:confiniily. ' massive Thfelding 'of
~
~
the. samp1'a... station.. and'separ,ati.on of.ttie. samp1e station and readout, in-.
. ;strumenation is. required for. accid.ent" condition. usage.
As the det;ectprs.
are sensitive to other sourcesof neutrons, location of the detectors within the plant should be considered, and the detectors should be lo-cated'away from these sources.
~
.1he sensitivity of baronometers is on the order of L pps baron with a useful rartge of. 5000 'ppe or more..
Calibration.of baronometers, can. be performed. stati-cally or by fTowing standards with a variety of baron concentrations past the detector.
Although no laboratory testing was performed,on the.
effects of sample matrices,. no chemical interferences are anticipated.
A
- - '....,. '. ^
The advantages of baron analysis by baronometry are the continuous readout of tho baron concentration,l.the ' wide measurement range with or without. sample dilutions, the applicability of the method
^
to routina and accident use, and existence oi proven baronomei:ers.
., The disadvantage of the method is the use of Targe
,.a-..f---
voTume sampTes', whic!r couTd creata maintena'nce prohTems shou.1d a failure occur during an accident.
However, the impact of such an occurrence could be minimized as a. backup boron analysis c::pability using grab samples. is required for inline monitoring methods.
As noted above, high radtation' fields will not affect the performance of boron analysis performed by baronometry.
4.2.1.5 Dicichem Analyzer of Manual Mannitoi Titrations.
The procedure for baron analysis using either the Digichem Analyzer or i
manual titrimetry methods is, in principle, the same.
Mannitol is added
.I to th's sample to form a baron mannitol complex; hydrochloric acid is added to initially idjust the pH of the solution to 4.4; and the sample
.-,--..,......--.-.._.S.-.
x.
7..;.. p ;._ w m.
j
.g
~is. titrated to the, and point (pH 8.5)1 with sodium hydroxide.
The boren
' content of.'the sample.is derived' from the volume of sodium hydroxide titrant used and.ccapariston to standards data.
~-
The difference in the two procedures is obvious; one employs hands-on techniques and the other employs remote analysis.
The remote analysir is performed automatically with the Digichem Ana-lyzer. It makes use of a microprocessor for sample and reagent dispens-ing., solution mixing., and concentration measurements.
The analyzer auto-
--l maticaTTy i:aTcuTates' th'e boron content 'and outputs it on a computer-compatible tape.
Analysis by the analyzer can be performed continuously, semicontiituously,' or in 'the batch mode. ' Separation"of the sensing ele-ment and the readout device is required to eliminate radiation effects on the system electronics; the sensor.and electronics can be separated ~
~
by at least twenty-five feet without degradation of the signal.The analy-
~
sis __ times are seven minutes with the automatic analyzer and twenty
~
minutes for the hands-on methods.
.~
N
~
I
[
A total of two hundred micrograms of baron is re-
' quired for Janalysis with /either thea automatic'or.manua0 procedures. Ac--
1
~
cordingTy, the required sampia s,1zes depend on the concentration in the sample. For example, under the assumption that 0.1 al of reactor coolant is an upper Timit for the reactor coolant sample size, the initiaT con-centration' of' boron in the rea0er 'cbolint wouTd have to be two thousand
^
pga'or greater"to. provide sufficient boren for anaTysis.
The two thou-sand pga represents the icwer Timit of detection for. 0.1 mi samples and, as a result, precludes the usage of the hands-on mannitol titration pro-cadure usage on accident condition sampler.
However, it doer not pie-ciude the use of the Digichem Analyzer for accident conditions as larger samples can be collected. and analyzed remotely.
t In fact, the Digichem Analyzer has been laboratory tested on standards, with and without the presence of potential inter-farences; accurate, precise,,interferenca free results were abtained.
The measurement range of the procedure / for a 4.0 m1 sample is 50-6000
(
ppn baron, which could be extended downward by the use of larger samples.
23
.1 : ' '.: ~.
+ -4C 7-.~;,-+;- M
..7.;
~..
- i...
.. m The advantage of the automatic ~ mannitol titration
)
' ' ' ,proMure is'fith relatSEsim' l.'idIty, remotfoperit3cn'a1ckactEdisijcsf ' ' "
~
p
. utility under routine : and.. accident. conditions,.and.yide, measurement
.... range.-
y The only apparent disadvantage of the procedure is the potential. maintenance difficulty whicit might occur during replacement of sensing elements under accident conditions or rapid repair of the
~
microprocessor. However, as backup capabilities to analyze baron samples *
.are-required for. Intine sampia methods, the Digichen analyzer should '
seet all measurement requirements.,
u.
The effects of high radiation fields have not been tested. ENICO feels that,the effects probably will not be significant; however, this shouTd be confirmed.
l 4.2.1.5 Ion Chromatography (IC). An icn chromatograph oper-y* _
ates arr the principle of selective' retention and eluti,on.of ionic species
- (
,on and from ion exchange ~ media.
It basically consists of a separator columrr and'eluents a suppressor. ceiumn. a conductimetric detector, and 'a
~
i readout dev. ice.. T.o perform an analysis for anions, such as horates or chTorides, the sample is fir.tt passed through the separator cahimn - an
~
an.iorr exchange medium. which retains the anions and re' places them with another anion from the exchange medium... The _ retained anions are then.
selectively removed from the seperator eniumn with the eiuent, normally a dfiute salt solution, and passed through the suppressor column. In the suppressor column - a/catiort exchange medium - the anions are converted l
to-their acid. forms, whictr pass unratarded. ta the conductimetric detector.
The conductivity of these diluta acid. solutions. is a. function of the
. anion concentrations irr the sample.
i The time betweert sample injection and the appearance.
~
of' conductivi,ty peak for a particular anion depends on'the sample size,
(
l l
the physical size of the columns, the types of exchange media,. and the
)
l types, concentrations, and flow rate of the eluent.
As a result differ-l ent anions in a single, sample can be separated and analyzed b'y proper selection of parameters.,
S--..--
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~
~
~
_.,.T T2 L.
..... a ga.. ;.
..v..
. c sin the developnent.of, an fon, chromatographic proca-dure for.the. analysis of boron and/or chloride; SEC/hus. studied various combinations. of.eiuents, separator columns,.<sup;tressor columns, and sam-
. Initial test' ng resulted 'in a method which.
.ple injection loop -sizes.
i used a. sodium tetraborat~e eluent and was applicabTe for' chloride analysis of postaccident reactor coolant samples,(cf Section 4.2.2.1).
- However, the analysis of boric acid salutions with the procedura showed inconsis-tant results.
~
- AdditionaT' deveTopnent and testing by Dionex, the a
manufactu-er' of the. ion chromotograph used, resulted in a procedure for the ' simultaneous ~ analysis' of baron and ' chloride using a single sample. i 1
'In,the test program a. modified Dionex Model 10 Icn Chromatograph was used. The modifications incTuded two 4 x 250 mm seoara-tor columns, a.3 x 250 ma suppressor co(umn,.'a. twenty c:n~(0.043 ml) sam-ple injection loop, and a sodTum carbonate / sodium hydroxide /mannital eluent.
An additonai ? requirement identified was the need of a cation p
p,re-column to remove excess base and convert bo' rates.to ' boric acid prior vtM' loading highly' basic samples..into ' bha 2injdctiori Icop.
With-a twenty--
~
five percent pump stroke, the necessary. times for the baron and chloride
~~
peaks te appear following injection to the sampling Toop are respective-ly E-6 and 2.-10 minutes. ~
'..9 ? -.d,.;..ug;.:. ?, y-r ; r; rd i, =. m.
z.c :-. -
'- v= :.5-
~-
~
s.
-\\
. c Th consistently obtain satisfactory results, peri-odic washing and/or regeneration of 'the suppressor and pre-columns is nece'ssary The pre-column requires. regeneration after the analysis of every-two. to. three samples containing 0.4. M sodium hydroxide.
The re-quired frequency for washing and regeneratierr of the supressor columns j
was not stated.
However,. hased on the frequency noted irr the inital chiaride analysis development work, estimated frequ'ency for regeneration is every four hours of continuous operation.
The need for this is indi-cated by an erratic baseline on the readout device.
The required fre-quen'cy for washing the suppressor is once daily or prior to each regeneration..
f 25
e.
- 4. - --.
~,,..
~
s rd
.j y,_
- 3..
s..
.y.:.
yIf column washing and regenera. tion are not. required,z
. r.-
the analysis ~ time'is forty minutes.
If coTumn washing and regeneration "are reg' ired' prior toi anaTysis', the sa' 'ple' analysis" time is ~approximateTy ~
~
~
u m
- two'. hours. ' Neither' case '# nclu' des ' system calibration ' time, which'.'is' i
fifteen minutes.
The IC procedure'for simult;aneous chiaride and baron analysis has been laboratory tested using simulated postaccident reactor coolant samples, stable fis,sion products, caustic, cooling water impuri-
' ties,' and normaT.Na'ctor coolant chemical additiives.
No sample 'niatrix l.
~
- ~~
. effects were observed within the specified. measurement range.'
The advantages of the procedure are its adaptability to remote operation,. the large chlorida meas'urement range, the simplicity
' of. operation, small sample. sizes, potentially short-sample analysis time, e
and the Tack of chemicaT interferences.-
A
~
The disadvantages 'of the procedure are the lack of a sufficient me.asurement range..for..boren, the need of a pre-column for basic samples, and the need for column washes and regeneration which -
~
might lead te Tong-analysis times.
~
The effects of Targe irradiations associated with highTy radioactive sampTes have not.been, evaTuated..' However,, based on a T,iterature study of radfation effects on the components of the IC and on limited laboratory tests used te determine the effects of 0-200 ppn.hy-drogen peroxide in samples,. no radioTogical effects are anticipated.
The Titerature showed that cation resins begin to ' degrade at' approxi-E mat:aly ID rads and. that the eTectronic ccmponents are resistant to 5
exposure welt above 10 rads Botfr levels are well above those anticipated to be encountered by the IC' during analysis of samples.
4.2.1. 7 Carminic Acid Saectrochotometry.
Two procedures were presented for boren analysis with carminic acid, one by SEC and one by GE.
The one presented by GE was detaiTed; it was developed by HACH 20 Chemical Company and closely "follows an ASTM procedure.21 The
<.... e,,.. :.,.
.,.N..._,.,~,,..,.,,,.....
- p., j..
,.e.
s procedure presented.by SEC'was only an. outline'.
S, ace.beth methods wiere.
similar and the HAQt procedure -had' a slightlf. Targer measurement range,
~
orily the HACH procedure is discussed.
The HACH procedure is very simple.
First Ihe carminic acid in preweighed tablet fann is added to 75 ml of sulfuric acid and mixed; then, 35 mi of the prepared solution is added to 2.0 mi of the sample, blank, and/or standard.
After the development of the 1
i color, 20-30 minutes 25 m1 of the solution (s) is transferred to spectro-
-l is.L i.ric ce1Tr. and the percent! transmittance is measured at 605 nm with a Bausch and 'Lomb Spectronic 20 s'pectrophotmeter,. or equivalent.
~
The measurement range is 0-15 ppa baron without sample dilution and 0 -
l several thousand ppm baron with sample dilution.
Tne total analysis time is aw. aiaiately 40 minutes -
..;. a The procedure has not been, tested for pcstaccident-reactor coolant sampla chemical matrix effects; it has been tested for
.3 effects of high sampTe radiation fieTds.. At the maximum anticipated 3
,. source. term, 8. x.10 rad /h for a 0.1 ml. reactor coolant sample diluted.
to '25.0'*mPths ' effects if trradiation should 'be" equivaient to 'no 'most.
i
- than 5 pps baron.; 7.This wouTd resuit in' negligihla error wheri, compared.
to invels of baron in postaccident samples.-
u:.::. m.::d.c.21The advantages of the procedure 'are the smaTT sample ~
~
sizes required. the wide measurement range, the adaptability to routine and accident conditions, and the simplicity.
The disadvantage-is. lack of laboratory testing with postaccident chemical matrix samples 4.2.1.8 Conductivity of Scron Solutions. A GE. specification-requires the Standby Liquid Control System (SLCS) at SWR's to be filled with a solution of borax and 3cric acid at a ratio of 1.028. ~GE proposed i
that, in the event the SLCS were actuated, the borort concentration in the reactor coolant could be estimated from conductivity.
GE tested the 27
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~
.....u_.,.4 q;....
....m.
~,..
)
....[..,. hypothesis.w.ith.a.1.028. borax. to. boric acid solution.by,. varying the.boren concentration between 5.4.and 201 ppa boren.. The. calibration curve was lindar between 10.8 and 201 ppn' borone.
n This ' suggests that, with' sample dilution, 'the 'boren.
~
concentr;: tion of reactor coolant can be determined by conductimetric measurenants -- However, ENICQ believes that under accident condition
~
there are too many other variables which could ' affect the conductivity of the reactor coolant and cause erroneous measurement.
Accordingly.
the ipproach is itoit' c6nsidered tot bei applicahTe for measurement of boron' concud. ration in reactor coolant.
4.2.1.9 Sumary and Conclusions for Baron Analvsis Procedures.
The results of ENICO.'s evaluation of potential chemical p1alysis procedures and methods for postaccident reac-tor coolant sample baron ' analysis' are sumarized irr Tab ~le 5.
Included are the measurement ranges, se5iittivities, accuracies, analysis times, '
sample sizes,. and anaTysis methods'. Also noted are the complexity.of'
~~
~
the procedures and the existence, based on actual testing and/or pro.fes-
- ' ' sional jud'gements, of known or ariti,cipated chemical or radiological in-tarfarences. - Fin *J1y,,.the applicahility of the procedure to routine and accident condition use is indicated.;.
^^
~
r t. :g: &
.k-As a71-but.one,of fthe. procedures.meth or exceeded.
c the criteria for required' sampTe size, radialogical exposures, measure-
~
i ment range and accuracy, and arialysis times;. the selection and ranking l
of the procedures in order of applicability were based to a degree on the complexity of the procedure and the _ laboratory testing which had been performed If two procedurer had similar complexities or amodnts of laboratory testing; other factors,. like time of analysis., were con-1 sidered.
Inline analysis procedures were ranked lower than grab sampling procedures with similar quaTifications as the capability ta analyze backup grah samples is required.for inline methods. Last, anticipated maintenance problems or potential contamination were considered.
3 i
1
_ _ _ _ _ _. _ _ _ _ _ _. _... _ - ~. _ _ _ _, _. _
_.-_._~._-_.,.-.a__.
,.,...y_,_,,._,_.,#-______.,
-, ~., _
i -~--
--- ;- - -,. ~.
.r
- e..
~..
TABLE 5
' FEATURES OF PRCPOSED. ANALYTICAL PROCEDURES FOR BORON Method Fluoreborate Curcumin
. Plasma Feature Electrcde.
Spectophotometric Spectroscopy RANGE (ppn)
Direct Analysis -
-G.5 ~f.G m -- 0~.2 - 2.G 0-<1.0
~-
With'1.:100 0ilution 50-600 20-200 0-<100 With 1:1000 Oilution.
500-6000 200-2000 0-<1000 With other Oilutions 50-6000 20-6000 0-6000
>t.ICG (2)
Accuracy (%)
+30 if 8 = 50 113
+20(3)-
~
(8 in ppn) 110 if 8 >300 p
Sample and/or Analysis Method?
Inline No No No Grab Yes Yes Yes
~
Analytical Backup Required?
No No No Sample Cbliection Method
~
)
Sample Analysis Method
- ggpffgg..g7). - -
U.
- Ofluted Apa3ysts SampTe 1.0-5.0
- 1.0 G.25 Actual RCW G.01-0.05 0.01 4.006 Analysis Time (min) 20 120 30 Precedure Ccaplexity Mediu:s ~ :... &.
- 1. m a - - =
Medim:r Low
~
demicaT Interferences?
No No Unknown Tested Yes Yes.
Limited Anticipated No Radialogicai Effects?
Unknowir Unknown Unknown Tested.
Na No No Anticipated No No No Apolication Routine Yes Yes Yes Acciden,t Yes Yes Yes A sa
[
C=7;
. ' '..~~....,._Z.Q.g G,.. :..g.a.f_,. g.,
C.l
.<.n
...... i;
..,w..
c TABLE 5 (Continued)
,s
. s.. c.,
...,.s.,
.. FEATURES OF PROPOSED ANALYTICA
L. PROCEDURE
S FOR BORON
.c.: n i
Mannital Titrimetry.
Method Carminic Acid (Manual 'or Digi-Ion Fcature Spectrophotometric Chem. Analyzer)
Baronometer Chromatography RANGE (ppa)
-Direct Analysis 0-10.0 50-6000 0-5000(1) 500-6000
=
. With.1:100 Dilution 10-100.
Not appropriate 0-500.,000 Not appropriate.
With 1:1000 Dilution 100-1000 due to Tack 0-5,000,000 due to lack of With' Other Dilutions.
.~0-6000 of.s'ensitivity.'
0-6000 sensitivity -
>1:100(2)
- e. -
Accuracy (*.)
+15(3) y
+g
+g B in ppa)
(Sample and/or Analysis Method?
Inline No Yes Yes.
Yes Grab Yes No No No Analyticai Backup:-
Yes, For f..
' Required?-
No-Yes, For Infine Inline-Ne r. - J' Samplo Collection Method
- Samp1s Analysis' Method
~ AvaiTable..
Available U
Not Specified-
~Not Specif.ied -.
E01Tuted AnaIpis Sample -
0.0Z
~ ^
4.3(E)'
Samp1'o Size (al).
2.0 1000-2000 O.04 1-2 10-20 0.04(5)
~
Actual RClW
~
AnnTysis Time (sin) 4Q(61-5-3G continuous40-120(7) x.a :%;
n.v.
n
-Procedura Complexity Low Low Medium Chemical Interferences?
Unknown No Unknown No Tested No Yes No Yes.
I Anticipated Na No
~
~
RadioTogicar EffectsT Nor Unknown No '
Unknown Tested Yes Na yer No.
Anticipated.
No No.
Application
.Yes.
Yes(8)
Yes No(9)
~
Routine Yes Yes Yes Nor Accident
, Notes:
- 1) The range of measurements using neutron adsorption is based on boron densitometers used at the Idaho National Engineering Laboratory.
1 30
.a.g-- amr
.m,
...s.
o.
1
(
measurement. range' ckn be extended to ten-of-thousands of
- 2) With dilutions. greater than,1:100 the. upper limit. of the"
~
pga. However an upper limit of 6000 ppn is noted as-measure-k ments.above 6000 ppe are not required.
f
- 3) In the~ procedure pra'sented the uncertainty of the method was not included; based on professional judgement the uncertainty has been estimated at +20 percent.
- 4) The actual. volume of reactor cooTant used in the analysis was determined from a 100-fold dilutic'n of 0.1 ml of reactor coolant and the volume of diluted sample required for the analysis.
.... ~
5} Due to a Tack of sensitivity for boron, typical sample dilu-tion of 1:100 of 0.1 m1 reactor coolant samples is. not appro-priate. Consequently, baron analysis of grab-samples can not.
be made with the procedure.
However, the procedure has suf-ficient sensitivity to analyze chloride, in diluted grab sam-pies (see sections 4.2.2.1 and 4.2.2.5).
. 6)~The procedures were presented for baron analysis with carminic acid, one by SEC and one by GE.
Thg analysis time specified by GE and 'SEC were 40 and'90 minutas, respectively.
The difference in tfbes is the number of minutes required
.for cooling following carminic acid addition and for coice Q~.
development.
As GE had' tasted the procedure and SEC had e
not, 40 minutes is assumed to be correct.
V.
- 7) Tfie actua'I' baron ahilysi's: time;is 'fehy (4d)' minuted. ' Mcw
~
ever, during continuous operation a column wash / regeneration /
equiTibration cycia is required every four hours. According-ly, an analysis could require approximately two hours.
?
- 5 8) The manual.mannitoi titrimetry is appropiate for routine usa only as the method Tacks sensitivity to analyze smali reactor coolant samples; ditions.the manuaT method is commonly used at pWR's under normal con The Dig 1 Chem Analyzer method is applicable to routine or accident condition usage as the method uses remota analysis of larger reactor coolant samples.
- 9) The. ion. chromatograptric procedure is not appropriata for routine or accident condition usage due to insufficient sen-sitivity If %e-lower detection.of 500 ppm baron were deemed to be sufficiently sensitive,. the procedure would be appropriata for accident condition use only w
9
/
C e
- l i
)
31
.. n. ;..,.,.:...,.:..>.. +...
f Listad, in order of appropriateness is the result of
'ENICQ's evaTua' tion 'of'the b'dro'n analysis ' procedures?
[
t 1.
Fluoroborate Electrode
. 2..Digiches Analyzer, Mannitol Titrimetry 3.
Ciarcumirr Spectrophotometric
- 4.. 8cronometer 5.
Carminic Acid Spectrophotometric 6.
Piasma Spectroscopy A. Ion Qiromatography It should be. emphasized..that the. order of ranking fr based on presently available information only.
With additional test-ing the order could change.
For example, with confirmation that there are no chemicaT intarferences to the canninic acid spectrophotometric method, it would be ranked at or near thc top due to ease of usa.
Like-wise, modifications to the plasma spectroscopy instrument; which t.ould
~
f.
insura containment. of volatilized radioactivity,.would. improve its rat-ing. Last, confirmation.of' the existance or nonexistance of radiological
~
' interferences could alter: the order of ranking.
i
~ 4.Z.Z ChToride AnaTysfs Methods ~~
^
4.2.2.1 Ion Chromatograchv (IC).
Described in Section 4.2.1.5 was art font. chromatographic.p.wedura for the simultaneous analy-u :..
- e...
......,, m - -. n.- -
sfr of baron and'chioride. ' Included in the description were the columns, sample sizes, eTuent, and operational characteristics required for satis-l factory analysis of boron and chloride.irr a single sample The. measurement range,. accuracy, sample: size and analysis time for cMorida analysis: wittr the procedure are respectively 0.12000 pps, t 10. percanti,. 0.04. mi of undiluted reactor coolant, and 40-120 minutes.
The procedure, witich has been laboratory tested, is applicable for routine and accident condition use.
It can also be used as an inline monitor or for analysis of grab samples.
e l
32
- a. l,...
.T
~ ~. ~ ~
~; ~
~ ~ '
~
.s.
c.
~ '
. The advantages of the procedure are the measurement rarige for chioride "normai 'and accident usage, small sample size, the
- i lack of chemical interferencak, remote ope.rability, simplicity of op' ra-tion, and potentially short analysis time.
e 1
The disadvantages of the procedure are the lack of a. sufficient. measurement range for.boren, the required column washes /
regenerations, which increase the anaTysis times, and the need of a pre-column for basic samples.
Another unknown is the lack of data on the potential. effects of highly radioactive samples..;
1
. 1 also~ developad :and,
. Sentry. Equipnent Corporation.
~.
tested another ion chromatographic procedure for chloride analysis.
The procedure can not be used for baron analysis; however, it is very similar to the boron-chToride analysisi described previously.
The procedure uses a 3 x 250 mm separator eciumn, a 6 x 250 mm suporessor eq1umn, a sodium tetraborate eiuent, and a 0.04. mi.. sample.
The ' procedure does, not use a
~
pre-column.
To c'htain satisfactory results the columns must be washed
' rioi-to each and regenerated.
Washing. frequency is once daily or I
p Y ;.. r e n e r ati.o n.-
3. 2...
Regeneration frequency is one every four hours.of cdntinuous. operation. A high erratic basetine,, a change in the time of
.the, appearance at the..chlorida, peak,(,normally,six minutes), and/or a
~
change in Ore pe'ald' height-for 'the standard ' ndicata e need for i
regeneration.
The tetraborate IC procedure has. been tested in the laboratory with simulated sampler of fission products. and. chemical addi-tives.
Speciai laboratory tests. were performed to determine the effects of morpholine,. hydrazine, amonia, and; natural and synthetic oils.
The i
only affect observed was due to sils,. which caused a. progressive 10-30 percent increase in the chloride response and a memory effect.
- However, as the memory effect can be eliminated with column washing and regenera-tion and as the increase in chloride, peak height is associated with
(
longer slution times, the effect is not considered.significant as it can be detected and corrected.
33
~
....._..2,
+.
6.
3......
!.aboratory tests were also performed dith the te-
' ' 'thabdatA IC procedure todet'armiNi it's.?abi1Ity to measude [ fluoride anif '"
)
.fodide.
The1 data. indicated that. the'- fluoride.elution time..was 1.,5
-minutes and that measurement of. fluoride 1,s possible down to 25 ppn (?10 percent) in. the presence of fewer than 100 ppa baron.
Attempts to mea.
sure fluoride in the presence of higher concentrations of baron. were unsuccessfhI due to peak overiap. The fadida measurements indicate that-
~ iodide could not be detected at low concentrations (0.5 ppn), and at high concentrations-(urto 100 ppn) smali responses were observed.
The fodida data indicates that. iodide will not. interfere with the tetraborate.
~
The advantages and disadvantages of the tetraborate IC procedure are essentiaTTy the same as.the ones presented above for the baron-chloride IC procedure.'
4.2.2.2 Scecific Icn Electrode (SIE).
The procedure for chloride analysis by SIE is very simpTe and rapid.
The,pH of the' solu-p tion is adjusted to 2-4 and the SIE and a reference electrode are im-mersed in the' solution aitd the millivoit " response is relate'dto t'he'chl'o.-
ride conc =J.rdtion.
~
The investigative studies performed by SEC/NUS em-pToyed a Graphic Controls Ilitra.-Sensitive SoTid State Chicrida Electrode (Madet PHI 91100) and'a Graphic Controls'doubTe-junction reference elec-l trade (No. GC 54473).
In the procedure 1.0 mi of nitric acid was added to 100 mi of sample to adjust the pH. The measurement range determined with standard. chloride salutions was. 0.01 to 35',000' ppr chloride.
With the; above measurement range, the SIE.is an-pTicable to routine use only-as approximately 10.0 mi of reactor coolant
~
sample, diluted to the 100' mi sample analysis size, would be required to detect 0.1 ppu chloride.
Furthermore, the method suffers, from inter-ference of other halogens.
The interference problem possibly can be solved by a combination of selective oxidation and solvent extractions; however, at present the SIE is not applicable to postaccident chloride
)
analysis due to the relatively large sample size required.
34 l
~
'L
_ -.-. ~......,,
- 1.. -
m
....-. r. ~.. e..
r.
?
Conceivably. the SIE could be adapted to remote h
operation; [.but, as noted,'. the. chemical. interference problem must be.
solved. Overall, the method is not a good candidate. -
y.
'4S2.2.3 Turbidime'tric. C'olorimetric. Titrimetric and
'S m.i.rvohotmetric.
' General Electric and SEC/NUS evaluated 1r suggested a nusber~ of other candidate procadures for chlo-
~
~
ride analysis.
ATT are basically hands-on methods; however, one (titri-metry) could be adapted to remote inline analysis.
There has been limited or no laboratory testing of the procedures by SEC/NUS or GE in 5
4
. regard to their applicability ta ad$1ysis of reactor coolant samples
^
~
with potential fission product,or, chemical interferences., However, based on the judgement of personnel at' ICPP who have 'pri,or experience with the same problems on similar procedures, it is anticipated that fodides and/
~
or other haTogens wi1T interfere with all the procedures presented in this section.
Futhermore,. due to the relatively large size reactor cool-ant samples required for analysis, 2-50 mT, use of the procedure's for hands-on analysis is prohibited under accident conditions., Accordingly, Q.
the procedures are not applicable'ta analysis of postaccident ' samples V. without further. testing,. modification, and developnent, or without remote
. For infomational
- purposes, each precedure is briefly 'outi fned below.,.,,,.,,.,,,,,
Turbidimetric and Colorimetric The turbidimetric and colorimetric procedures are very similar.
Six dropr of concentrated nitric acid are added to the sample,12 al for cotlorimetric and 25 m1 for turbidimetric; the percent transmittance ir recorded, seven drops of' I N silver nitrate. are added; and the precent transmittance is recorded again.
The difference between the two recorded measurements is related to the concentration of chloride by the use of calibration standards.
For turbidimetry a HACH Turbidi-meter or equivalent is recomended, and for colorimetry a Coleman Nepha-Colorimeter, or equivalent, is recomended.
(
35 ha 9
~-
..w.
,... ~
.. ~
... c.
e
\\.
.'Soectrochotometric t.
simple. and... commonly used for., chloride. analysis...It.. involves the mixing of.10 ml of. ferric annonium sulf.ata. solution, E.0 ml of mercuric thio-cyanate methanol solution, and 25 mi of sample.
Tais is followed by the measurement of the percent transmittance at 463 nm in a 10 cm spectro-photometric celt.
Titrimetry TFre. titrimetry method is based _ on. the.formatica.of a mercury complex, diphenylcarbozone-bromphenol
- blue, and mecurous chloride. The e:id-point color develop::ent occurs whenmercurous. ions are
~
in excess of the chloride.
In the procedure 25 ml of sample,1-2 ml of diphenyl-carbozone, and a few drops of the bromphenol blue indicator are mixed. This is followed by the addition of mecuric nitrate.
The quanity of mercuric nitrate added is a function of the chloride. concentration.
2 4.2.2.4 Conductivity 6f ChToride ~!alutions. 'For.a dilute e
so.lution of an ionic s.pecies the specific conductance, X, in pS/cm is
given by:.
3 K = 10 AC (4-1.)
where 'A. is. the~ equivaTent conductivity 'and.C is the concentration of the.fonic species in salution in electrolytic equi-valents. When the conductivity of a seTution is due to seieral fonic species,. the specific conductance of the solution can be expressed as the su::mation of the conductances of each of the separate ionic species:
'K 101 [,, ( A C )
(4-2) g where A and C are respectively the limiting j
j equivalent ionic conductance and conceritration of the individuaTz species in solution. Values, whichr are available in handbooks, of th_e equivalent conductance of different ionic species can be used to calculate the con-ductivity or, altern'ately, the concentration of the ionic species pro-I vided the ionic species concentrations are known or the conductivity of
~
the solution is known.
' 36
'i
~
. 2,-.c -
- ?.
.....,[...,...
....s, The proposed procedure utilizes the above techrIique for esdmation o'f' upper'liarits of 1 hloride' concentratifoni in. postaccident reactor. coolant samples.. ENICO agrees,such a'tachnique is applicable'for
. estimation of upper, Timits of chloride or.other. ionic species in solu-
~
tion, but does. not believe the technique meets the intent of the NRC re-guirement for chloride analysis.
For exaniple, chloride concentrations calculated front the conductivity of postaccident, solution will, in til probability, he in excess of the 0.1 ppn limitation due to the presence of fission products, high *adiat. ion fields, and/or other chemicals.
As
.a result, s. e.uve-actions. will. be; taken. or,, more likely, accurata analysis of chloride concentrations will be made.
Initial accurata de-terminations will preclude undue enneern and/or unnecessary actions.
4.2.2.5 Sumary and Conclusion.of Chloride Anaivsis Procedurer. At present there is only one applicable method for chloride analysis of postaccident chloride. analysi s; ion
. chromatography. 'The other procedures evaluated are not appropriata due ta the Targe sample sizes required and knowr or anticipated chemical 1
interferences t,o the procedures.
The results and featurns of the proce-
.dures evaluated. are shown,.in Table 6.
- ' ' The chemical procedures have not been ranked in order of applicabilitiy.
Of the methods not presently applicabTe, the specific fan electrode' and the titrimetry methods appear to have the most potentiai due ta adaptabil'ity to remote use, i.e., reductierr of radiclogicai exposures.
Their use, however, will depend on eliminatierr of~ chemical interferences,, such as other halogens.
Limited investigative work was performed 'by SEC'/NUS to eliminate the chemical inte'rferences.
Their technique, which.ENICO' believes has good. potential, was' selective oxidatierr - solvent extractierr.
ConsequentTy,. with additional testing and developnent one or more of these procedures could be adapted for
~ '
postaccident use.
Specific procedures, proposed in the future will re-quire evaluation as they become available.
4.2.3 Dissolved Hydrocen and Oxycen r
(
4.2.3.1 Gas chromatocraohy (GC) - Hydrocen Analysis.'
A gas chromatograph consists of a sample injection loop, a chrematographic 37
.-_~n,._
c 7
. r....j.. -
.,,.. w o
column containing a media'such as charcoal or molecular sieves, a thermal "cienduct'ivitica'TI, 'anii'a" mets -re'adout ' device.
The iherirar~cond' ctivity' '
)
u cell. or.detecto'r has(.two,. matched hot ' wire filaments.
Two streams of.
carrier gas, e.g., argon, are supplied. to the GC from a comon ~ source.,
' ' One stream flows directly past one of th'e filameni:s; the other stream flows through the GC. calumn then to the second hot wire filament.
In the absence of a samille, the two filaments. reach thermal equilibrium (constant resistance) and no detector output is observed.
Upon injectson J
of a sample into the GC column, non-equilibrium between the two filaments
_l
- is. created.dua ta.the different. thermal.conductiv.ities of.th.a gase.s j
.eTuted. from the GC calumn ta the sample stream 1ilament.
Tha thermal conductivity imbalance generates a detector output.
~l As the different consti.tuents of a sample are eluted from'the GC column at different and' specific times, the observed detector outputs can be attributed to the individual component of the gas sample.
The magnitude of the outputs are..proportionai 'to the concentrations of the different gases'in the sampleeQuantification of the concentrations f-..
is achieved by comparisic.n of the detector output of samples and (L) standards.
The GC suggested by GE is, a Baseline ModeT 1030, or equivaTent.
The Modet 1030 is a microprocessor controlled instrument witiv thermal.. conductivity detectors.
It is equipped with a gas condi-tiener, an automatic retention time. indicator, and thermai conductivity peak integrator The suggested GC column is ten feet of 1/a to 3/16 inch tubing with SA malecutar sieves. The carrier gas (helium) flowrate,
3 and pressure are 30 cm / minute and.15-30 psig; the suggested column temperature was 30-50"C Although a F1 sher-Model 1200 Gas Chromatograph war used in the SEC/NUS developnent and testing program, SEC/NUS also sug-gests Baselin's GC, or equivalent,, for plah ' applications due to its
~
larger measurement range. Specific GC columns and operational par'ameters were not given by SEC/NUS.
It is assumed the specifications will be similar to those noted by GE.
Many coinbinations of columns, carrier gas flowrates, and temperatures have been used successfully in the past.
38
r,....
~ ~.
6.:
=.
,. ~....
TABLE 6
' FEATURES OF PiOPOSED ANALYTICAL PROCEDUREI FOR CHLORIDE-s a
Specific
. Method
?-
"" ' Ion
~'
~
Ion Feature Chromatography Electrode Turbidimetric
' RANGE (ppa) lysis-
~
~ Direct Ana 0.1 - 100 0.010 - 35,000 0.02 - 10 10 - 10 000 Not Applicable Not Applicable With 1:100.011ution O.I h,CDU due to lack due to Tack -
overaTT Range-of sensitivity of sensitivity _
.u,.
~
(C1 in ppa)
+15(1)
+ 20
+ 30(2)-
Sample and/or
~.
Analysir Method Inline Yes Yes No Grah Yes-Yes Yes Analytical Backup oquired?
Yes, For Yes, For
. Nol..
CJ1aple Collection Inline Infine(4)
Available Available.
Method
~
Sampic. Analysis Method F".
Not Specifind '-
Not Specifiedt
./
..'.5 Samo10 Sfze (mi)-
DiTuted Analysis Sample 4.04(6) 100(7) 25(7)
Actual RC
' 4.04 10 5
, Analysis Time (min}
~
'40
.120{81
' 15,
20 Procedure Complexity Medium Low Low 6 amical Interferences?
No Unknown Unknown Tested Yer Limited No Anticipatad Yes Yes Radiological Effects!
Unknown Unknown Yes(9)
Yes T sted No No Anticipated Ne No Application I
Routine Yes Yes Yes No-No Accident Yes -
e 39 J
-u
...4
...x....
- c. ;
TABLE.6.(C,ontinued) e.
..... c
.s...
)
FEATURES OF PROPOSED ANALYTICAL PROCEDURES FOR CHLORIDE' Colorimetric Ti.trimetric
,5pectropnotometric s.
RANGE (ppm)
. Direct Analysis.
With 1:100 Dilution - '.
0. 0 4 - 111.
0.1 - 10 0.02.-10.(1
~
Not Applicable ' Not Applicable Not Applicable.
Overall Range
'due to lack due to lack due to lack of sensitivity
, of sensitivity of sensitivity
. L,_gy (g}.
i:.
.. q.w;
~p.,,,.
. ~.
.+25(2)-
+20(3).
t20(3)
(CI. in ppe).
' Sample ahd/or
~
No Yes' No Analysis Method Yes
. Yes Yes '
Diline kd Analytical Backup No Yes, For) tio i
In11ne(5 Required 7-Sample Collection Available
-- Mathod Not Specifiert Sample AnaTysis Method
~
~
Sample Size'(al)
- 12.
100(8) 25(7)
Diluted Analysis Sample
- Z.4(7)'
50 '
5 Actual RC.
^
Analysis' Time (min) 30..
20 20 Procedure Complexity Law Low Low c.
Chemical" Interferences 2 '^
Unkitower
Unknown Unknown Tested Na No No Anticipatad Yes Yes No
~
Radiological Effects?
Unknown-Unknown Unknown i
Tested Na Na
'No Anticipated.'
Yes.
No Yes j
Application No No No Yes No Yes Routine Accident.
~
j Notes:
~....
- 1) The accuracy of the IC measurements is + 15% in the 0.1 to 1.0 ppm chloride range and is 125% for higher concentra-tions. By calibration at higher concentrations, the accuracy
}
can be maintained at t 15%.
s 1
40 i
.s u-
- 2) The uncertainties were estimated from calibration curve data presented:in.the associsted documentation.
(
The uneartainties are based on.profes'sionalljudgement. '
3)
~
- 4).The SIE methodccould be.. adapted for inline use.
- 5) The titrimetry procedure could be used as the inline method by employment of a technique similar to the Digichem Analyzer method for. baron.ana. lysis.
- 6) The ion chromatographic procedure uses small
(< 0.4 ml) undiluted reactor coolant samples.
7)' 'Due cte ' insufficient sensitivity. smaller reactor coolant samples' are inappropri. ate for these methods.
The' t' trimetry procedure has ' sufficient sensitivity to mea-
'8) i
~
sure 0.1 ppa chloride; however, 50 m1 of reactor coolant is required.
The method is now in usa at LCFT at INEL.
g) Limitad radiological effect' testing was perfor=ed by GE on the turbidimetric procedure.
At the maximum anticioated dose rate, 8 x'103 rad /h in a diluted 25 ml samole (0.1 mi diluted to 25 ml), -an equivalent resconse of 1.3 pcm of chloride. was calculated from measurement cata.
.-}
l
..'v
.,. ~
~
- i..'
e 9
h^
O a.
t 41
.......m.
.. _ _...._...... _... _ ~
,. y.
,.m..,...
The sample collection procedures. for dissolved gases
('.' 'Ipfoposed 'by SEC/Nb!"and Gfar limiTaN The;GEp@ediid'ureihvo.Nis' th'e' iso-
)'
.Tation of 70 al.of pressurized reactor, water, the. depressurization of.the sample into a 20,al gas holding con,tainer, and.,the transferral.of ali-quots from the gas. holding container to 15 mi septum bottles. The 15 ml septum bottles are transferred to the. laboratory for GC analy' sis and/ or thrther dilution.
In the Taboratory, gas tight ' syringes are used to take 1.0' mi aliquots from the septum bottles for injection into the GC.
1he procedure employs Henry's Law and a tracer gas, which is injected j
fato the.. sample priar ta depressurization ta daterm.ine.sampia yield..
y
. The.. sample collection. procedure of SEC/NUS involves s
the isolation of a 30 mi pressurized reactor water sampia, depressuriza-tion of the sample, the quantitative transferral of the dissolved gases into's 300 mi gas hcTding cyTinder via an argart gas purge, and the pres-surization of the 300 mi cylinder to atmospheric pressure, with the argon j
purge ga'.
From the 300 m1. gas. holding cylinder small samples, 0.25 or.
s
--_- 1.Q m1 are injected.remotaly into.the GC for m alysis.
e-.
\\.
G Following. collection of the, di' solved gas samples s
the time requi
.for GC analysis is less thart ten minutes.
The measurement range reported by.'SEC/NUS is based on extensive
- Taboratory studies and. is appitcabTe for 25-25,000 ppm hy-drogen for a'I'0 si dissoTved gas sampTe.
The'dtlutions associated with the sample, collection procedure and the 30 mi sample used for depres-3 surization create a range of 0.5 - 2000 cm of dissolved hydrogen per k.11ograr of reactor coolant The accuracy of the measurements is t 10 percent.
General Electric did not report a measurement range; however, an estimata of the lower limit of detection was mentioned.
Their estimate of the lower detection limit, based on limited laboratory
~
studies, is 0.1 volume percent or 1000 ppe for a 1.0 ml dissolved gas sample.
ENICO believes. this detection limit is a factor of ten or more high and that the actual detection limit will be similar to the one reea-
)
sured by SEC7NUS;-he.,100 ppm or lower.
If such a detection limit is verified by GE, ENICO estimates the measurement range of the GE gas 42 L
a
_.g_. _.
.,,., s,
chromatograph method for dis. solved hydrogen in reactor water will be 3 -
" ' '. 41-2000 cm. per kilogram of water." Tlie estimate is' tiased 'ori the
\\
~
. relative. size of reactor water, samples taken for analysis and the rel.a-tive volume,s of. the gas. holding cylinders.of the SEC/NUS and GE -sample collection systems.
..g.
,Tha, advantages of_ the GC methods proposed by the two vendors are sufficient measurement ranges, the application to rou-tine and accident usage, the simplicity of operation, and the selective measurement of hydrogerr.. not total gases. The advantages of the SEC/NUS,
method over the GE' method are the extensiveness of laboratory testing performed by.SEC/NUS -and the remote analysis capability of the SEC/NUS 14 system.
The lattar advantage is quite significant as calculated dose rates due to noble gases associated with the dissolved gases ii.
unit volumes of reactor water are in excass of 104 'R/h 'at one centi-meter.
As a result the dose rates associated with the GC. samples, even with d.1Tution, are potentially a--few R/h and w'ill require more caution for hands-on analysis.than rs' nota analysis.
., i !.. '.
.,..IIIe. d,isadvantage of,'the GC method.in. gene'ral, is,,
' related ta maintanance of the instrument; however, this is not considered significant 'as GC's 'are generaTTy very dependabis.
Another.11mitatiert
'of 'the method is the lack of Taboratory tests on the effects of high radiation field.. on, the.. procedure; however,,there. are no anticipated etfacts.
?
4.Z.3.2 Gar Chromatocrachy. Yellow Scrinos Analyzer -
Dissolved Oxygen Analysis As described irt the previous section,. different. constituents in a gas sample are separated l
in a. GC columrt due ta their characteristic diffusion ratas. througtr a medius such as charcoal,' molecular sieve,. etc.
As a result GC lands.
~
itself to the simultaneous determination of oxygen and hydrogen from the j
analysis of a single sample.
j
= =..
General electric proposed to use this technique for
(
dissolved oxygen analysis,1.d., simultaneous measurement of hydrogen and oxygen in a single sample.
The sample collection procedure, instru-mentation, and associated equipnent proposed,are identical to the ones described above.
Specific measurement ranges were not provided by GE.
43
,n i
. ;. v....... -
6..;..
.... x v.....,
.r.,.....
..a,...,, y.
g ENICO's estimat3 of 'the mea'surement range is #1
- t. c...
.to 400 ppe in the reactor coola...,.. Th'a b'a'si s of.. +..,.,...the estimate are the-
.. +...
c.
nt
)
j
. helative thermal.. conductivities 2Z '-('de' tactor. responses) of.. oxygen [and.,,
hydrogen: and the hydrog'en. measurement range. estimated, f.or.the GE system.
in the above section.
This estimated oxygen.is inadequate for postacci-dent analysis of reactor coolant samples as the sensitivity of the pro-cedure is jesufficient ta measure.below 1 pga dissolved' oxygen.
How'ever, before the GC procedure is precluded from postaccident application, it should be experimentally verified that the sensitivity of the GC method
.is inadequate.. : -:.m. u (s.
Sentry ~ Equipment Corporation proposed ~ an inline monitor for postaccident determinations of dissolved oxygen in reactor coolant.
The instrument selected and Taboratory tested was a Yellow Springs' Instrument (YSI) Model' 54 ' Oxygen nalyzer.
The sensing probe, which contains a. semipermeable membrane, is remotely located frem the meter and output device.
The probe holder was redesigned to minimize
~
fluid volume and associated redtattorr expo'sure' Calibration. of the system is achieved with an oxygen saturated dui.ineralized water source.
5 The
- actual' oxygen content of the.' standard solut'io'n is determined fror.t the temperatures of the water and a soTubfif ty. chart relating dissolved
~
l oxygen to water temperature.
i Lahoretary tests' verified that there were no inter-forences 'due to hydrogen in solution or ' variations in sample flowrate.'
One probles observed during the. tasts was a pin hole in one of the probe mostranes. Thir resulted in erratic results.
Replacement of the mem-brane corTseted the problem Laboratory testing also verified that the accuracy
, (t,55) was sufficient to measure 0.I ppa dissolved oxygen The measure-
~
ment range was linear betweert 0.I-7.85 ppe oxygen.
Concentrations above '
7.85 pps oxygen were not laboratory tested.
It is anticipated the mea-surement range will be valid up to 20 ppn oxygen; however; this needs verification.
)
Provided the measurement range can be demonstrated to be 0.1-20 ppm, the YSI Analyzer is appitcable to postaccident 44 6
a
.w
.?~:,,., L.,
.. ~..,
I applic tions.
Th3 Model 54 Analyzer lacks, sufficient sensitivity for routine use'.
SEC/tiUS ' proposed. a Rexnerd i Analyzer.for. routine ~ use (sansitivity-ppb). or alternately a Model 56. YSI Analyzer with reported higher sen'sitivity.
The routine condition monitor will be installed in
~
~
7 parallel with the accident conditian ~ monitor.
~
~,
The advantages of.the YSI exygen monitors are the
~
simp 1' city, and potentially adequate measurement remote operabflity, i
range.
.l The disadvantage of the s'ystems is the time, 1-4 hours, required.for the system ;to reach equilibrium. after the internal portions of the sensing probe are exposed to air.
Based on a review of I
the literature on the ef#ects of irradiation on the components of the
~
sensing probe, no radiological effects.are anticipa*:ed.
The maximum.
dose. anticipated to the different materials of constru: tion in the probe
~
is 10 rads; the minimum doses causing damage to the materials was O
6 reported at 10 rads.,
.:. a..,
]V
' 4.2.3.3, _ Evaluation Sumary of 01ssolved Oxveen and Hydrocen knalysis.
Thle gas chromatographic methods proposed by SEC/ flus for dis-solved hydrogen analysis ist appItcable to postaccident sa=ple analysis.
It has sufficient sensitivity, accuracy, and range of. measurement I
(0.S-2000 cn, hydrogen per kilogram of reactor coolant, t,10 *.).
The measurement' range assocfated withi the GE method needs to be verified.
After completing collection of the dissolved gas sample, the analysis time is ten artnutes or less.
The SEC/NUS sample handling has an advan-tage over the GE procedure due to its remote made of sample handling.
Precautions should. be taken when manually. handling the dissolved gas samples due to the. potentially higit radiological fields., The GC method is applicable to routine conditicas also Radiological interferences to
~
the GC method are not anticipated.'
.3 The GC method proposed by GE for dissolved oxygen analysis appears to lack sufficienti' sensitivity for required measurement of low '(<0.1 ppn) concentrations of ' oxygen.
Without further testing t
to demonstrate the capability of the method to measure the low concen-tration, the GC method for oxygen analysis is not applicable to post accident sample analysis.
e.-
r.
.. :.....,.. n..
s..
n.....
+
c.
e The inline oxygen monitor proposed by'SEC is appit-l cable"to' posticcilienDan'alysis yr6vided'jnformatiod'1s' ava'fiible ' tB" J!
verify,its ability.to measur.e dissolved. cxygen over.the entire range of.
0.1-20 ppa. At present the measurement. range has. been demonstrated.to.
~
be valid between 0.l'an'd I.85 ppn only.
In ENICO's opinion, the proven measurement range is sufficient as the intent is to measure the absence of oxygen, not neces.-
(
sarily the presence.
If NRC does not agree, additional laboratory
. studies need to be perforsned ta extend the measurement range.
l To meet all NUREG-0737 requirements, licensees which use inline method for analy'is must have a backup capability to obtain s
grah samples and 'to perform analysis performed by the inline monitor.
01ssoIved oxygen analysis 'by hands-orr' techiques will require diTuted,
~
pressurized samples or techniquer to collect the gases foxygen) frem a Ifquid sample.
Conventionai methods, eg., Winkler, of hands-on 'alialysis
.can not be used due to the large sample.sfzes required and/or a lack of sensitivity. Alternative methods must,be identified.
t.2 4 Conductivity and pH 4.2.4.I ConductivitN Both 'SEC/Nd5 an'd E propose the use of inline monitors for measurement of conductiv'ity.
Their proposed con-
....-500 pS/cm and 0-100 ductivity meters ' have measurement ~ ranges of 0
(,,,
us/c:r,' respectiveTy.3".The proposed probes have,~ conductivity ceTis with 0.1 as celi constants; they are located remote to the meters.
The inline probe tasted by E was a standard, commerciaTTy available one, and the -
probe. tasted. by NUS used a. modified probe holder designed to minimize fluid. volume The actual celt volumes were not spec.ified.
The ac:uracy associated. with the measurements was not specified either; however, higtr accuracy for 0-E uS/cm and decreasing accuracy for higher conductivi-ties was noted.
l l
Laboratory tests were performed by both GE and SEC/
NUS. The GE tests involved measurements of the conductivity of water flowing first through a conductivity' celf under irradiation and then
)
I through a second in. series conductivity cell not under irradiation.
The 4
5 irradiation fields were varied between 1.3 x 10 rads /h to 9.8 x 10 46
[
4.*. < *
.,......g...
3,.
rads /h. E also made static.(no flow) measurements with the above ar-
, rangement. ' Finally. [EE' performed conductivity measdraments on 'a 10 pan,'
^
~
~
r q
'. chloride, solution' win (9.'8' x 105 7,gsg),,fd without 1rradiation' and,' -
.with and without. flow through.the cells. The SEC/NUS laboratory testing was timited,to establishing the operability with a flowing sample stream, the effects air bubbles in the air stream, and the effects of hydrogen peroxide on conductivity measurements.
SEC/NUS also conducted a litera-ture review for potential radiation effects to the components of the sensing probe.
. 4.c
...,y The results of E tests on the two in-series cells indicated that the cell.under irradiation and the one not under, frradia.
tion gave the same results anEf that the conductivity of the sol'ution increased frem 0.'l.uS/cm to 0.65 US/c:: as the irradiatien intensity i
was increased.
The'cause for the increase in conductivity is unknown;-
however, it obviously is due to the generation of an unktwwn. conductive species.
The hypothesis that the. unknown species is hydrogen peroxide is not supported by.the chloride'salution tests performed by E and the.
i p
hydrogen peroxide tests performed by SEC/NUS; f.e., the conductivity did
,. not. change, with.,the', addition. of, chloride
.added,to, decrease,the., gene rated species - or with the addition of hydrogen peroxide directly to the fTowing streanr. -
'-'"?-
l i
'.: M t:. Further results of the SEC/NUS tests show that the monitor is appifcahTe to a flowing sample stream and that the presence i
of air bubbles at five percent of the water volume does not alter the accuracy of the measurements.
l The.11tarature review indicates that the resistance l
of the probe components to radiation exposure exceeds the anticipated.
radiation does by a factor of one hundred or more.
i The advantages of the method are its utility under
' " accident and normat condttion; f ts' resistance, to ' radiation damage,' the remote operational mode, and its simplicity.
k j
There are no apparent disadvantages even though the i
conductivity of water solutions increased with increasing radiation
)
doses.
This observation only implies that the monitor was operating i
47 4
=
~
^
. ~. u r.-
.y
...a ~. a.
...v....,.
.a.
property as its response increased with an increase in the conductivity
,q,yg y;,ggg,.
r....v..
- 3.,.: ->.
~-.
.(.:
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}-
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A backup. method.for measurement of, conductivity of
.~
grab samples,was.not noted. However, there are copnercial,1y available -
portable conductivity meters which are ' appropriate for this purpose.
- n. _
4.2.4.2 g. To investigate' methods of determining pH under postaccident conditions, SEC/NUS evaluated an industrial grade inline pH probe and a sealed. perunnently-filled reference electrode. The, vertical,
i l,
proba holder, modified.to minimize fluid volume,. prevents entrapment of 4
. 41r bubbias. A double G-ring seal is used' to prevent Teakage. The probe i
can be calibrated in place by injection of buffer' solutions (pH 7 and
- 10) into the sample loop.
The probe output.is recorded' on an industrial l
i i
gnde meter mounted in a remotely Tocated instrument panel.
1 i
Testing of,the pH monitor' was performed to deter-
] '
. eine its applicahtitty.to a. flowing. sample stream..and to avaluate 'the
(
c ff l
el ects of air bubbles in the liquid.
Data indicated the pH monitor is
.not-affected by variations in_ flow or by the presence of air bubbles.
\\
t
.c :.
I The optimum operating tamperature range of the pH 8
8 j
probe is. 75-90 7'; the maximum tamperature and pressure are 125 F and 10E paif. lHth constant. control of the pH probe at a given temperature f
, ithin the optimus operationer range, the accuracy and measurement range -
w i
will compiy with NRC requirements, 1-13 1 0.3 pH units.
1 As inline pH monitors have been used reif ahly for a l
nuder of years in the. chemical and, nuclear industries,. there. are no j
apparent disadvantages The advantages are 'the. application to accident f
and routine use. the remote operability, the simplicity, and the suffi-t I
'~
cient measurement range and accuracy.
{
(
To fulf111 411 NUREG-0737 requiremeits, huwever, a I
backup capahitity to measure the rH of grab sample mu.t be provided.
i ifith the proposed grab sasple collection systems, this will. require pH I
)
1 measurements of diluted reactor coolant samples.
ENICO anticipates that pH's measured in diluted samples cannot be used to accurately determ'ine l
j U
c_
a
...m
. A,**,... 4 7.
n...
...,e the actuai pH of the reactor water. For exampTe, based on a 1:100 diTu-
'~
tiidn of,e"0.1 mi reactor coolant s. ample,'wtth deionized water- (pH 7.0)',. '
th' estimetad pH$of tho' reactor cdolant. wa.ter. determined free.the an_aly-e
. sis..of the d1Tuted sample could be in cerror by 2-7 pH un,its..This.tak'es,
into account the results of dilution only and not the presence of. other constituents which can affect the pH.
Acco'rdingly, ENICO does not re-commend the use of dfiuted grab samples for. measurement of reactor cooTant pK.
3
.. General ETectric Company. suggested the. usa.of pH
-a paper for measurement of pH in. reactor water.
In can, junction with this tdea, a seies of laboratory tests were. performed to determine. the ef-l facts of high irradiations on the accuracy of the method.
The pH paper
!I.
was famersed in solutions with pH 3.8. and 10.0' and irradiated for ten 5
4 einutes- (I.f' * '10 " rads) in one study and one minute (1.5 x 10 rads) in another study.
The colors of the solutions..were c:mpletely
' destroyed in the ten minute test and significant1r t".ared, in the one minute test (0.5 pH unit shift). U q
v
- 3... Ja.. compensate,fon fthis.affact, l.GF., suggested, that 1.
the,.he be modified ta decrease the exposures to the pH pacer.
The suggestion was to sofsten ?.he pa'oer with a drop or two of' samole instead of totai fumersion of the pH paper in the sample solutico.
The proposed modified, Jare was not demonstrated ta be successfui.
1
.3
,,, 3
..,...s ENICD does not beffeve'the pH paper method is sat-1sfactory at present due to the irradiation effects observed.
Its future applicability will depend or the additional testing and the development of a technique ta callect a smell, undfluted reactor coolant sample and to perfors'the measurement in a, radioTogicaT1y safe manner.
j Senerat Electric' also suggested that the conducti.
~
vity of a solution is a potential method to ensure that th,e pH,of tb
~
reactor casiant fr'wfthin 'certain ' acceptable ranges, i.e.
5.5 to 8.6.
j ENICO does not believe this technique meets the MAC' intent as the con-
[.
ductivity of the reactor coolant can' possibly vary over a large range under accident conditions and cannot be used as an indication of pH with the required accuracy.
49 F
.,..n
.y e.,5 e*
-*s e.
'.45 *
- o s
- s ', ~ **
}
,. [.
o e
a 4.2.423 Sumary of Conductivity and pH Analysis Methods.
" The imethod.jiropos'ad'by SEC/NOS and GE is " applicable' for' measurement of-
[
~
i
... - the conductivity of.re. actor coolant water. ~.,It inclu fes ar.inline con-
.ductivity celi.with a. remote readout. meter.
A backup capability to mea-
.sure the conductivity.cf grab samples was not noteii by '5EC or GE.
This
~
backup capability can be provided with commercially available, portable conductivity meters for analysis of grab samples. Alternataly, a backup capability,. with NRC's concurrence, wou,1d be a second, independent inline monitor which could be put into service upon the failure of the first monitor.. The monitors,are applicable to accident and normal conditions.
ATthough there was an. increase in conductivity 'of test solutions with an increase in radiation exposures, the effect was not due to monitor component failure.. It was a result of an increase of the. conductivity of the test solution. It is unknown if the increase in conductivity of the test solutions was inherent to the eperimental con-ditions or whether one should anticipate the generation of a conductive species tre postaccident re' actor coolant water.
,. m
)
!.At.present the only proven methods which. has satis -
factory accuracy and is applicable to measurement of pH under accident conditions ara intine' monitors.
The use of pH paper is not applicable due to inaccuraefes caused by high radiation fields-The pH paper method may be applicahTe with further testing; however the. method will require the deveTopnent of techniques for the remote ad'dition of smail reactor coolant samples to the pH paper and for the remote comparison of the pH paper colors with standards.
The pH analysis of diluted reactor coolant samoles is not recomended.due. ta the potential inaccuracies of the measurements.
The only alternative for a backup analysis capability is the use of two in(ependent inline monitors; one in service and one in standby.
The pH inline monitors are applicable.to both acci-dent and routine use. There are no anticipated radiological effects.
i 50 x,.
i g.W.
- 7..
~.
.j.,.,
3....-
- 7.., J.
.o......
...,....n.
0 5.0 REFERENGS p'
- 1. THI-2 Lessons Learned Task Force Status Report and Short-Term Recem -
mandations, NUREG-0578, July 1979.;.
/
~
- 2. Eisenhut, D.
G., letter to All Operating ' Nuclear. Plants, 13 Septeder 1979.'
- 3. Denton, H.
R., Letter to all Operating Nuclear Power Plants, 30 October 1979.
l
- 4. NRC Action Plan DeveToped as a Result of TMI-2 Accident, NUREG-0660, l
May 1980, revised August 1980.
- 5. Eisenhut D. E.
Latter to All Licensees of' Operating Plants and
/
Applicants for* Doerating Licenses and Holders of Construction-
, Permits,. f.5epted er 1980.
~
~
~
- 6. Clarification of TMI Action Pian Requirements, 'NUREG-0737, Novehe'r; 1980.
- 7. McCracken C., Telecopy to R. Huchton, Initiatien of Work on Project
- 3 (Post Accident Sampling), 23 April 1981.
S. Priviata Commupications, C. McCracken to R. Huchton,13 August,1981.
j r-lm,
- 9. Emel W. A.
Letter to C. E. Etimore,, " Proposed Work, Schedula",
WAE.I8-81, il May 1981.
4
- 10. Instrumentation for Light. Water. cooled Nuclear Plants to Assess Plant J and' En' irons Con'ditions During and Pollow'ing an' Accident, Regulatory
- v l
, Guide 1.97. Revision 2. Oscader 1920.
~ 11. Assumptions Ussd for Evaluating the Potential Radiological Consequen-ces of a Loss of Coolant Accident for Boiling Water Reactors, Regu-latory Gaida 1.3, Revision 2, June 1974.
p.
- 12. Assumptions Used for Evaluating the Potential Radictogical Conse-quences of a. Loss of Coolant Accident for Pressurized Water Reactors, Regulatory Guide 1.4, Revision 2,, June 1974.
- 13. GDC 19, Appendix A 10 CFR Part 50 14 Heldoir, L R.
Post Accident sample Station Activity Source Terms, General Electric Nuclear Engineering Division document ORF 000 3, March 1981.
- 15. Developnent of Procedures and Analysis Methods for Post Accident Reactor Coolant.
Nuclear Utility Service Corporation (SEC), April 1981.
- 16. Ta'chnical Description and General Information, General Electric Company, NEDC.24889, Section 1, date unknown.
51
_. V Af\\
1~
~. -,...
..'. o,,.
.s, 2 4... <.
.-,.....,./
2 La f 1
- s.. ; 7,. Operation. General., Electric' Company, NEDC-24889 Section 6, dato,
. g m ;...... -
g.
a.
).:
18.' ChemicaT/Radi.ochemical. Procedures.... General. Electr.ic Company.. DRF 0004, Section 7 and Appendix A date unknown.
'Is. SWR Generic Post-Accident Samplin Design Requirements,.
General Electric Company, C5474-SP-1,g. SystemRevision 1. April, 1980.
- 20. Soron Carmine Method MACH Chemical Company, APHA Standard Method, 14th Ed., p. 2so, 1s75.,
- 21. Standard Test Method for Boron Analysis in Water, Method A - Carminic j
Acid Colorimetric. Method. AST.M 03082-74.,
a.,,
1 22.
R.- H.
and Chilton, C.H.,
Editors.
Chemical Engineers!
, Fifth Edition, McGraw-Hill Book. Company, p.,3-215.,1973.
~..
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POSTXCTGEEM sv. FLING SYSULP.
NUREG-0737, !!.B.3 EVALUA110N CRITERIA GUIDELINES
~
'The post (ccident samoling system will be eval at d for compliance with u
e the criteria from NUREG-0737 II.B.3.
These eleven items have been copied verbatim from NUREG-0737.
The licensees submittal should include information equivalent to that which is normally provided in an FSAR.
System schematics with sufficient infermation to verify flow paths should be included, consistent with documentation requirements in NUREG-0737, with appropriate discussion so that the reviewer can determine whether the criteria have been met.
Further information pertaining to the specific clarifications of HUREG-0737 consiccred in the reviewers evaluation are listed below., which will be Technically justified alternatives to these criteria will be considered.
itericn:
(1) The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples.
The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sample.
arification:
Provide information on sampling (s) and analytical laboratories locations including a discussion'of relative elevations, distances and methods for sample transport.
Responses to this item should also include'a discussion of sample recirculation, sample handling and analytical times to demonstrate that the three-hour time limit will be met (see (6) below relative to radiation exposure).
Also describe provisions for sampling during loss of off-site power (i.e. designate an alternative backup power source, not necessarily
- 0L the vital (Class IE) bus, that can be energized 4n sufficient time 4,jf to meet the three-hour sampling and analysis time limit).
itYrian:
(2) The licensee shall establish an onsite radiological and. chemical analysis capability to provide, within 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; iodines and cesiums, and non-
~
volatile isotopes);
(b) hydrogen levels in the containment atmosphere; dissolved gases (e.g., H ), chloride (time allotted for (c) 2 analysis subject to discussion below), and boron concentration of liquids.
(d) Alternatively, heve inline monitoring capabilities to perform all or part cf the above analyses.
i i
e e
e e
e s
Clarification: 2 (a.) A discussion of the counting! equipment capabilities is needed, t
including provisions to handle samples and reduce' background radiation to minimize personnel radiation exposures (ALARA).
Also a procedure is required fo-relating radionuclide concentrations to core damage. The procedure should include:
1.
Monitoring for short and long lived volatile and non volatile radionuclides such as 133Xe 131. 137 s 1
C 134Cs, 85xr,140 a, and 88ge (See Vol. II, Part 2,.
B pp. 524-527 of Rogovin Report for further information).
2.
Provisions to estimate the extent of core damage based on radionuclide concentrations and taking into considera-tion other physical parameters su.ch as core temperature
~
data and sample location.
2 (b) Show a capability to obtain a grab sample, transport and analyze for hydrogen.
2 (c) Discuss the capabilities to sample an'd analyze for the accident sample species listed here and in Regulatory Guide
.d 9l * *v :.2:....
~7
'2 (d) Provide a discussion of the reliability and maintenance ' -
~
information to demonstrate that the selected on-line instrument is appropriate for this application.
(See (8) a and (10) below relative to back-up grab sample capability Q
and instrument range and accuracy).
r M iterion:
(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.
~C1arification:
System schematics and discussions should clearly demonstrate that post accident sampling, including recirculation, from each-sample source is possible without use of an isolated auxiliary system.
It should be verified that valves which are not accessible after an accident are environmentally qualified for the conditions in which they must operate.
Criterion:
(4) pressurized reactor coolant. samples are not required if the licensee 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 ade::uate. Measuring the 0 concentra-2 tion is recomended, but.is not mandatory.
Clarification:
Discuss the method whereby total dissolved gas or hydrogen and oxygen can be measured and related to reactor coolant system concentrations. Additionally, if chlorides exceed 0.15 ppm, verification that dissolved oxygen is less than.
0.1 ppn is necessary. Verificatien that dissolved oxfgen is
<0.1 ppm by measurement of a dissolved hydrogen residual of 3
3,.._
~_~'"*q_
o
.{
.i
> 10 cc/kg is acceptable for up tc 30 days after the accident. Within 30 days, consistent with minimizing personnel radiation exposures (ALARA), direct monitoring for dissolved oxygen is recommended.
~
Iriterion:
(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 primary containment systems and the cooling water. Under both of the above conditions the licensee 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 licensee shall provide for the analysis to be completed within 4 days. The chloride analysis does not have to,be done onsite.
- 1arification:
BWR's on sea or brackish water sites, and plants which use sea or brackfish water in essential heat exchangers (e.g.
shutdown cooling) that have only single barrier protection between the reactor coolant are required to analyze chloride
~
- ~ ithin 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sE All other plants have 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to perfom
~ -"
w a chlorida-analysis. Samples diluted by up to a factor of one thousand are acceptable as initial scoping analysis for chloride, provided (1).the results are reported' as ppm C1 (the licensee should establish this value; the number in Q
tha blank should be no greater than 10.0 ppm C1) in the reactor coolant system and (2) that dissolved oxygen can be verified E
at <0,1 ppm, consistent with the guidelines above in clariff-dation no. 4 Additionally, if chloride analysis is performed
-~
on a diluted sample, an undiluted sample need aise be'taken and ret.ained for analysis within 30 days, consiste1t with ALARA.
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) (i.e.15 rem whole body, 75 rem extremities).
(Note that the design and operational review criterion was changed from the operational linits of 10 CFR Part 20 (NUREG-0578) to the GDC.19 criterion (October 30, 1979 letter from H. R. Denton to all licensees).
Clarification:
Consistent with Regulatory Guide 1.3 or 1.4 source terms, provide information on the predicted personnel exposures based on person-motion for sampling, transport and analysis of all required parameters.
Criterion:
(7)
The analysis of primary coolant samples for boron f s required for PWP.s.
(Note that Rev. 2 of Regulatory Guide 1.97 specifies the need for primary coolant boron analysis capability at BWR plants).
4 e
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f:larification:
PWR's need to perfom boron analysis. The guidelines for h
BWR's are to have the ca) ability to perform boro'n analysis but they do not have to c o so uni,ess boron was injected.
Criterion:
(8)
If inline monitoring in used for any sampling and analy- '
tical capability specified herein, the licensee shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing the samplies.
Established planning for analysis at offsite facilities is acceptable.
Eq0ipment provided 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.
Clarification:
A capab'ility to obtain both. diluted and undiluted backup samples is required. Provisions to flush inline monitors to facilitate access for repair is desirable.. If an off-site laboratory is to be relied on for the backup analysis, an explanation of the capability to ship and obtain analysis for one sample per week thereafter until accident condition no longer exists should be provided.
.... Criterion:
(9)
The licensee's radiological and chemical sample analysis
. =....
capability shall include piovisions to:
(a)
Identify and quantify the isotopes of the nucifde categories discussed above to levels corresponding to the c
.1
. source terms given in Regulatory Gui'de 1.3 or 1.4 and 1.7.
d-Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduc-tion of personnel exposure should be provided. Sensi-tivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concen-tration in the range from approximately 19 Cf/g to 10 C1/g.
(b)
Restrict background levels of radiation in the radiolog-ical and chemical analysis 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 use of a ventilation system design which will control the presence of airborne radioactivity.
Clarification:'(9) (a)' Provide a discussion of the predicted activity in the samples to, be taken and the methods of handling / dilution that will be employed to reduce the activity sufficiently to perform the required analysis.
Discuss the range of radionuclide'concen,
tration which can be analyzed for, including an assessment of, the amount of overlap between post accident and normal sampling capabilities.
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p (9).(b) State the predicted background radiation levels in the counting room, including the contribution from s~amples which
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are present. Also provide data pemonstrating what the background radiation levels and radiation effect will be on a sample being counted to assure an accuracy within a factor of 2.
Critei ton:'
(10)
Accuracy, range, and sensitivity shall be adequate to. provide pertinent data to the operator in order to describe radiolo-gical and chemical status of the reactor coolant systems.
Clarification:
The recommended ranges for the. required accident sample analyses are given in Regulatory Guide 1.97, Rev. 2.
The necessary accuracy within the recomended ranges are as follows:
- Gross activity, gamma ~ spectrum: measured to estimate core damage, these analyses should be accurate within a factor of two across the entire range.
i
,Boren: measure to verify shutdown margin.
In general this analysis should be accurate within +5% of the measured value (i.e. at 6,000 ppm B the tolerance is
+ 300 ppm while at 1,000 ppm B the tolerance is 150 ppm).
Tor concentrations below 1,000 ppm the tolerance band should c.,
remain at i 50 ppm.
- Chloride: measured to determine coolant corrosion potential.
For concentrations between 0.5 and 20.0 ppm chloride the analysis should be accurate within + 10% of the measured value. At concentrations below 0.5 ppm the tolerance band remains at 1 0.05 ppm.
. Hydrogen or Total Gas: monitored to estimate core degrada-tion and corrosion potential of the cool nt.
An accuracy of + 10% is desirable between 50 and 2000 cc/kg '
- but i 20% can be acceptable.
For concentration below 50 cc/kg the tolerance remains at i 5.0 cc/kg.
- Oxygen: monitored to as'sess coolant corrosion potential.
For concentrations between 0.5 and 20.0 ppm oxygen the analysis should be accurate within + 10% of the measured value. At concentrations below 0.5 ppm the tolerance band remains at 1 0.05 ppm.
+
e s
.s
- pH: measured to assess coolant, corrosion potential.
Between a pH of 5 to 9, the reading should be accurate within +0.3 pH units.
For all other ranges + 0.5 pH units is acceptable.
To demonstrate that the selected procedures and instrumentation will achieve the above listed accuracies, it is necessary to provide infomation demonstrating their applicability in the post accident water chemistry and radiation environment. This can be accomplished by performing tests utilizing the standard test matrix provided below or by providing evidence that the selected procedure or instrument has been used successfully in a similar environment.
STANDARD TEST MATRIX FOR UNDILUTED REACTOR COOLANT SAMPLES.IN A POST-ACCIDENT ENVIRONMENT
. Nominal Concentration (pom)
Added as (chemical salt)
Constitutent..
IF 40 Potassium Iodide Cs+
250 Cesium Nitrate Ba+2 10 Barium Nitrate La+3 5
_ i Ce+4 5
Ammonium Cerium Nitrate s/-
C1-10 B
2000 Boric Acid Li+
2 Lithium Hydroxide 150 M0'f NH 5
K+
20 Gamma Radiation 104 Rad /gm of Adsorbed Dose (Induced Field)
- )-
NOTES:
- 1) l Instrumentation and procedures which are applicable to diluted samples only, should be tested with an equally diluted chemical test matrix.
. The induced radiition.envircrrnent should be adjusted commensurate with the weight of actual reactor coolant in the sample being tested.
2)
For PWRs, procedures which may be affected by spray additive chemicals must be tested in both the standard test matrix plus appropriate spray additives. Both procedures (with and without spray additives) are required i
to be available.
3)
For SWRs, if procedures are verified with boron in the test matrix, they do not have to be tested without boron.
...~:: -..:.. :... - -
.i 7-4 y.
4)
In lieu of conducting tests utilizing the standard test matrix for instruments and procedures, provide evidence that the selected instrument or procedure has been used successfully in a sinilar environment.
All equipment and procedures which are used for post accident sampling and analyses should be calibrated or tested at a frequency which will ensure, to a high degree of reliability, that it will be available if required. Operators should receive initial and refresher training in post accident sampling, analysis and transport. A minimum frequency for the above efforts is considered to be every six months if indicated by testing. These provisions should be submitted in revised Technical Specifications in accordance with Enclosure 1 of NUREG-0737. The staff will provide model Technical Specificat, ions at a later date.
Criterion:
(11)'
In the design of the post accident sampling and analysis capability, consideration should be.given to the following items:
(i) Provisions for purging sam ~ple 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 ifmit reactor coolant loss from a rupture of the sample line. The post ej accident reactor coolant and containm1nt atmosphere samples a
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 system.
(b) The ventilation exhaust from the sampling station should be filtered with charcoal absorbers and high-efficiency particulate air (HEPA) filters.
Clarification: (11)(a) A description of the provisions which address each of the items in clarification 11.a should be provided. Such items, as heat tracing and purge velocities, should be addressed. To demonstrate that samples are representative of core conditions a discussion of mixing, both short and long term, is needed.
If a given sample location can be rendered inaccurate due to the accident (i.e. sampling from a hot or cold leg loop which may have a steam or gas pocket) describe the backup sampling capabilities or address the maximum time that this condition can exist.
BWR's should specifically address samples which are taken from the core shroud area and demonstrate how they are repre-sentative of core conditions.
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8-I, Passive flow restrictors in the sample lines may be replaced by redundant, environmentally qua'lified, remotely operated isolation valves to limit potential leakage from sampling-lines. The automatic containment isolation valves should close on containment isolation or safety injection signals.
(11)(b)
A' dedicated sample station filtration system is not required, provided a positive exhaust exists which is subseq.uently routed through charcoal absorbers and HEPA filters.
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POST-ACCIDENT SAMPLING GUIDE FOR PREPARATION OF A PROCEDURE TO ESTIMATE CORE DAMAGE The major issue remaining to complete our evaluation of NTOL's for compliance with the post-accident sampling criteria of NUREG-0737 is preparation of procedures for relating radionuclide concentrations to core damage.
To date, none of the applicants has been successful in providing an acceptable procedure. As a consequence, each NTOL has a license condition which may restrict power operations.
One of the contributing factors in the applicant's slow responses to this item is their confusion on exactly what to prepare.
The attachment is intended to provide infomal guidance to each NTOL applicant so that their-
- ' procedures, when prepared, will address the core damage estimation in a manner acceptable to us.
We anticipate that preparation of a final procedure for estimating core damage may take approximately 12 months.
Therefore, we are willing to accept an fnterim procedure which focuses on fewer radionuclides than are indicated in the attachment.
The interim procedure in conjunction.with a firm date for the final procedure would be used to remove the power restricting license condition.
F The primary purpose in preparing a crocedure for relating radionuclide concentrations to core damage is to be able to provide a realistic estimate of core damage. We are primarily interested in being able to differentiate between four major fuel conditons; no damage, cladding failures, fuel over-heating and core me t.
Estimates of core damage should be as realistic as l
possible.
If a core actually has one percent cladding failures, we do not want a prediction of fifty percent core melt or vice versa; extremes in either direction could significantly alter the actions taken to recover from an accident. Therefore, the procedure for estimating core damace
~
4_
should include not only the measurement of specific radionuclides but a weighted assessment of their meaning based on all available olant indicators.
The following discussion.is intended to provide general guidance pertaining to the factors which should be considered in preparing a procedure for estinating core damage but is not intended to provide an all inclusive plant specific list.
l l
The rationale for selecting specific radionuclides to aerforn " core damage estimates from fission product release" is included in the Rogovin Report (page 524 through 527, attached).
Basically, the Ronovin Recort states that three major factors saast be considered when attenoting to estimate core damage based on radionuclide concentrations.
1.
For the measured radionuclides, what percent of the total-available activity is released (i.e. is only gao activity released, is sufficient activity released to predict fuel overheating or is the cuantity of activity released, only available through core melt?)
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What radionuclides are not present (i.e. some radionuclides will, in all probability, not be released unless fuel overheatino or melt occurs).
The absence of these species bounds the maximum extent of fuel damage.
.3.
What are the ratios. of various radionuclide species (k.a. the gap activity ratio for various radionuclides.may differ from the. ratio inthepellet). The measurement of a specific ratio will then indicate whether the activity released' came from the gap or fuel overheating / melt.
In addition to the radionuclide measurements, other plant indicators may be available which can aid in estimating core damage.
These include incore temperature indicators, total quantity of hydrogen released from zirconium degradation 'and containment radiation monitors.
When orovidino an estimate of core damaos the information available from all indications should be factored into the final estimate (i.e. if the incore temerature indicators show fuel overheat and the radionuclide concentrations indicate no damage then a recheck of both indications should be perfonned).
O' Consistent with the categorization of fuel damage in the Rogovin Renort, h
the four major categories of fuel damage can be further broken down, similar to the following list, consistent with. state-of-the-art technoloay.
The suggested categories of fuel damage are intended solely to address fuel integrity for post-accident sainpling and do not pertain to meetino normal off-site doses as a consequence of fuel failures.
1.
No fuel damage.
2.
Cladding failures (<10".).
3.
Intermediate cladding failures (10% - 50%).
l 4.
Major cladding failures (>50%).
5.
Fuel pellet overheating (<10%)
6.
Intennediate fuel pellet overheating (10% - 50%).
7.
Major fuel pellet overheatino (>50%).
8.
Fuel pellet malting (<10%).
- 9., Intermediate fuel oellet meltina (105 - 50%).
- 10. 11ajor fuel pellet melting (>50%).
i e..
=
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8ecause core degradation will in all probability not take place unifonnly, the final categories will not be clear cut, as are the ten listed above.
Therefore, the preparation of a core damage estimate should be an iterative process where the first detennination is to find which of the four major categories is indicated (for illustrative purposes, only radionuclide concentrations will be considered in the following example, but as indicated above, the plant specific procedure should include input from other plant indicators). Then proceed to narrow down the estimate based on all available data and knowledge of how the plant systems fu,nction.
Examole In a given accident condition, there is 70% clad failure, significant fuel overheating and one fuel bundle melted.
Utilizing the iterative '
orecess First calculate the maximum fuel melted by arbitrarily attributing all activity to fuel melt (under these conditions, five to ten melted bundles may be predicted)., Therefore, the worst possible condition is fuel pellet melting."
Second, calculate the maximum fuel overheated, by arbitrarily attributing all activity to fuel pellet overheating (under these conditions, mjor fuel pellet overheating is predicted).
Third, calculate the miximum cladding failures, by arbitrarily attributing all activity to cladding failures (under these conditions, greater than 100% fuel cladding damage is predicted).
At this point is is obvious that major cladding damage is present and that a
~
large amount of fuel pellet overheating has occurred with the potential for some minor fuel pellet mel'ino.
Fourth, check for the presence of radionuclides which are indicators of fuel pellet melting and overheating.
In this instance, obvious indicators of overheating will exist along with trace indicators of potential pellet melt.
Fifth, based on the radionuclide indicators of fuel cellet overheatina damage (confirmed by incore temperature) make an estimate of how mch fuel overheated. This result will in all probability indicate
. mjor fuel pellet overheating.
Sixth, subtract the activity estimated from fuel cellet overheating,
~
olus the activity attributable to 100% gap release from the total activity found. This will result in a negative number because the contributions from overestimating cladding damage (100% versus 70%)
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and fuel overheating (major versus intermediate will exceed the activity contribution from one melted bundle.
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b At this point, knowledceable judgment must be employed to establish the best estimate olf core camage. Although all damage could be attributable to cladding damage and fuel pellet overheating, the trace of radionuclide indicators of fuel pellet melt indicate the possibility of some fuel melting. Based on knowledoe of core temperature variations, it is highly unlikely that 100% cladding damage would exist without significant fuel melting. Also, some of the activity attributed to fuel pellet overheating must be associated with the amount of fuel pellet melting which is indicated.. Therefore, the best estimate of fuel damage would be that " intermediate fuel overheating had occurred, with ma.ior cladding damage and the possibility of minor fuel pellet meltins in one or two fuel bundles out of 150 fuel bundles."
The above example is obviously ideal and makes the major assumptions that:
~
A.
The radionuclide/s monitored are at ecual concentrations in all fuel rods.
In actuality, at no time will all radionuclides be at equal concen-trations in all fuel rods. Because the time to reach equilibrium A
for each radionuclide is different, due to their hichly variable production and different decay rates.
Some isotopes will aoproach equilibriin quickly, while others never reach equilibrium.
Therefore, it is necessary to factor in reactor power history when detemining which radionuclide is optimum for monitoring in a given accident condition. Probably the optimum radionuclides for estimating core damage will vary as a function of time after refueling and based on o
power history.
B.
Equilibrated samoles are readily available from all samole locations at the instant of samoling. Considering the large volumes of O_
liquid and vapor spaces that a leakage source migrates to and mixes with, for other than very large leaks, it will take many hours or even days to approach equilibrium conditions at all samole locations.
C.
Maximum core degradation occurred orior to initiation of samolinn.
Unless total cooling is lost, core degraoation can be anticioated to progress over a period of hours. Thus, there is not a given instant when samoling can be conducted with positive assurance that maximum degradation has occurred.
Considering that ideal conditions will not exist, then procedure for estimating core damage should be prepared in a manner that the effects of variables such as time in core life and type of accident are accounted for. Therefore, the procedure for estimating core damage should include the detemination of-both short and long lived gaseous and non-volatile radionuclides along with ratios for appropriate species.
Each separate radionuclide analyzed, along with predicted ratios of selected radionuclides would be used to estimate core damage. This process will result in four separate estimates of core damage, (short and long-lived, gaseous and non-volatile species) which can be weighed, based on power history, to
. detemine the best estimate of core damage.
k -
The post-accident sampling system locations for liquid and gaseous samoles varies for each plant.
To obtain the most accurate assessment of core damage, it is necessary to sample and analyze radionuclides from each of these locations (reactor coolant, containment atmosphere, containment sumps and suppression pool), then relate the measured concentrations to the total curies for each radionuclide at each samole location.
These measured radionuclide concentrations need to be decry corrected to the estimated time of core damage (to).
Their relationcip to core damage can be obtained by cogaring the total e,uantity and ratio 3 of the radionuclides released with the predetermined radionuclide concentrations hnd ratios which are available in the core based on pows.r history. Assuming one hour per samole location to recirculate, obtain and analy:e a sample from each location it.would take hours to perfonn each of those analyses.
Based on the above rationale, the final procedure for estimating core damage using measured radionuclide concentrations will probably rely only on one or two samole locations during the initial phases of an-accident. The optimum radionuclides for estimating core damace will also, in the short term, be based on recent power history.
uhen eouilibrium conditions are established at all samole locations, radionuclide analysis The soecific can be performed to obtain a better estimate of core damage.
radionuclides to be analyzed under equilibrium conditions may be different.
')
2 than.those initially analyzed because of initial abundances and different decay rates.
The specific sample locations to be used during the initial phases of an accident should be selected based on the type of accident in procress (i.e. for a BWR, a small liquid line break in the crimary containment would release only small quantities of volatile species to the dry well.
Therefore, sampling the dry well first would not indicate the true magnitude of core damage).
For the same small break accident, if pressure is reduced.
by venting safety valves to the suopression pool, then the suooressionIn the chamber vapor space would contain the majority of gaseous activity.
case of a small steam line break, without venting safety valves to the suppression pool, the dry well may be the best samole location.
To account for the variations in prime samole locations, based on type of accident, the procedure should include a list of primary samnle locations.
This list should include both a prime licuid and gaseous location and state Additionally, the reasoning used to determine that these locations are best.
the procedure should address other plant indications which can be used to verify that the sample locations selected are best for the specified accident condition.
Finally, the procedure should incorporate olant specific examples which show estimates of core damage based on predicted radionuclide concentrations.
Methodology for this step is provided by letter of May 4,1981, from McGuire Nuclear Station, Docket No. 50-369.
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y at 2 h: :. 54 n: ntes, utncagn t.; J.t.c.:.:
t.
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stumc..; c: curred at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 45 minutes.
caloy in t.e cc:
- 7. At of the fuel roo3 in the cero burst, dunng an
- 14. The c: maga in 1:.. cc c c ti :s frcm the tcp accroximately 30-minute (center bund!e) to downward at less 7 fe t, rne probably 8 feet.
- 40. minute (fowest power peripheral bundles) over mest of the c:re.-d c: s'sts of oxygen pericd after the top of the core was uncovered embrittled Zircaioy c! add ng tec;ed by a bed of at depths ranging from 1% feet (center bund!e) debris that probab!y c:rcs s of fuel pa!!st frag-to 2 feet (peripheral bund!c) from the top of the ments, partially disse!ved fuel pe!!ets, she!!s of fuel rods.
Zircaloy oxide, and segments of embrittled Zir-
- 8. Temperatures at which Equefied fuel (UO dis-caloy cladding with outer skins of Zirca!oy ox-2 solved in the zirconium metal-ztrecruum dioxide ide, a!! glued together with Equefied fuel into a Squid eutectic at about 3500 to 36007) could relatively tight and compact mass extendng
~
be formed were calculated to have first been entirely across the core from wall to wad and reached at 6 inches from the top of the fuel in
. penetrated by only a few vertical passageways.
the fuel rods in the central fuel bunde about 33 at most. In addition, fingers of Equefied fuel ex-minutes after the top of the ccre was un-tend downward from the debris bed in several covered and were reached as low as 36 inches continuous subchannels between fuel rods, en-from the tcp of the fuel. Such temperatures compassing the neighbcring fuel reds, to a were calculated to have been reached in the depth of abcut I foct above the bottom cf the peripheral bund'es at a depth of about 14 fuel stack in the fuel roda. Not less than 32%
inches frem the top of the fuel in About 46 of the fuel assemblies have such fingers of minutes after the core was uncovered and at a lique'ied. fuel.
depth of about 41 inches ~in 57 minutes.
- 9. The peak temperatures calculated-for the fuel rods rar.ged from 43707 in about 52 minutes
- c. Core Ocmage Estimates from Fission O
for the highest powered bunde to a maximum Product Release of 44127 for a medum pcwered bundle at 58 minutes to about 43587 for a lower powered At shutdewn the reacter core contained fission pencheral bund!e at about 78 minutes.
products, activation products, and actinides. Scme
- 10. The amount of hydrogen formed by oxidatien of of these, notably krypton and xenon, are gaseous soGd Zirca!cy cfaddng durmg the temperature and can diffuse through the fuel pe!!st to cci!ect in excursion was calculated to be about 308 the gao between the fuel and the c!adcing. To a pounds, and that formed from all of the dam-lesser extent, the halogens (iodne and bromjne) can aged Z'rca!cy, inc!udng that contamed in the a!so diffuse into ther fue!-clad gap. Any perforation.
Equefied fuel present at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, was calculated of the claddng can release these fission products "
to be about 720 pounds. This is the rrtrumum into the reactor coolant.
amount of hydrogen estimated to have been If the fuel temperatures are higher than operating formed. The maximum cou!d be as high as 820 temperatures, but well below melting, other radioac-pounds.
tive materials are volati5 zed and can ciffuse out.
- 11. The maior releases of hydrogen to the contain-Also, diffusion of tha noble gases and halogens in-ment occurred before 4 hcurs accident *irne creases so that a larger fraction cf these can be and during the long depressurization around 8 released. The release of cesium is quite variable hours. No s!gnificant amount of hydrogen was and could be caused by compound formatien. Se-produced after about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
cause of this variability and what is now known
- 12. The minimum water level occurnng in the core about cesium, it is not possible to determine pre-up to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> is mtimated to have been 4t %
cisely the temperature at which a reasonably targe ft from the bottom of the fuelin the fuel rods on fraction of the cesium would be released; however.
the basis of the arr:ount of hydtcqcn produced, it is befieved temoeratures wculd not be lower than the amount of radioactivity re! eased, the time at 1300*C (2370'F). ar,ma "
which significant levels of radicactivity were At higher temperatures that cause the Ecuefac-
' detected, and the structural dzmage" estimated tion er melting of fuel, some fraction of cther fission
' in the core.
prcducts such as te!!urium can be released. Data k'
- 13. The total amount of Zirca!oy oxidized is calcu-reported show that the escape cf te!!urium depends.
lated to be not less than 16400 pounds and on many facters other than temperature." Under may have been as high as 18 700 pounds; i.e.,
oxidizing conditions some ruthenium may be 524
y_____
7 s
rs!Insed W:re me:t.ng. la ge: et.f. r.:/c Ltge the most sevarc'y c.-'agu-:f fuel. Sma fract::ns.
actions cf both te!!crium and rta."sn!t.m are approximately 10 % cr less. cou!d have been
.eleased in melting; but und r soma cend;tions, released frem perferated but other.,ise undamaged thess materials can also be released before melt.
rods, but this cannet te weil estimated.
The presence of ruthenium and tenurium does not prove that melt has occurred, but the absence of them is a goed indicator that melt has not occurred.
Leaching frem irradiated Fuel Moro recent experimental work,mr.so while tending Very small fractions of the remaining groups may to conAnn Amycus data has notfesolved aH me have been re! eased from the very hottest fuel. 'The 1
principal mechanism for release of these refractory
~~~~
W matet als is probably teaching. Laaching from irradi-i ated UO has not been ghly studied. Howev.
2
% W Sm Reelon produ' cts and most W h er, me,wm* W =ayama and W r-sym and
,,,,,,,,,,,,,,,,,,,,,,,y
,,,e,,,,,,,
Eklund, has shown that the teaching rates are raisesed only 'n W smal amomm m d D
elevated temperatures. However, if damaged fuel data, especiaDy for the temperatures and conditions peNets we W some W d's a %
existing in TM-2, are too sparse for a reliable cal-g ggg pg culation of the rate of teaching, especia!!y when one is&W@WkhWh matenais find their way into the coolant by teaching.
c ntact 4 water, tion plication is presented becausc geh g
the effectrve surface area of irradiated fuel present-ed to the water is almost irnpossible to estirna:e be-cause of craciung and porosity. 'The most that can Categor!es of Fission Product Releases and Their he done with the available data is to fcrm an *edu-Relation.to TM-2 d products and aednides can be M. cated guess
- as to whether the fuel accear p.
mha2 W very Ane Wts. Mout aenal data a a
{
which they are volatilized. One such grouping (from not posjar e to esdmate me acW size dstnbdm Ref.191) is in order of decreasing volatility.
of the av
.m... However, a sma5 fraction of the s
most refractory material can. e expected to have b
i Noble gases (Kr.Xe) found its way into the reactor coc! ant. An approxi-I Halogens (I,Br) mate teaching calculation is presented in Accendix 5 Alkal metals (Cs, Rb)
E.7. On the basis of mis acproximate calculation, it N TeGurbm (Te)
V Akaline earths (Sr. Ba) is possible to state, with very low confidence, that a VI Noble metals (Ru, Rh, Pd, Mo, Tc) large fraction at the fuel can presently be fragment.
ed and that the size of the fragments is more tikely VI Rare earths and we.
VII Refractory oxides of Zr and Nb to be a few mi:Etneters than dust 5ke. A similar cal-culation has been carried out by Pcwers.* His h fracum W p W volatRe h pro-c ncfusens, althcush not identical with these, indi-WW m h tem and h cate mat me obsM a% may have been-size of the fuel fragments, if the temperature is high D '***
- 8'*
- D or if me fuelis highly fragmented, neerfy We;;
monts or by distnbution of particle sizes no more man W um M matWals can k mmW.
than a few percent smal'er than 2 mil!imeters,in d,-i Under tu conditions mat have ban calculated ameter and none smaner than 0.6 ms!Imater in ciam-W h & W T M me g g,
release of groupe I and a can be assunnd from as fuel that was severely damaged. plus scme add-dami W M fuel mda h M., was Expected Discersion of the Fiasion Products from perforated without damage to the fuel. This add-N hw tional amount from perforated bid omstwise undam-Principal fuel damage probabty started before 3 aged rods is, probably partly belanced by me' hours after turbme trip. There was probably only
(
amount not released from severely damaged fuel.
A major fraction of group til and a much smaffer minor da.-. age before 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The calculated total invente,gs f fission products, activation products, o
fraction of group N could have been released from o
$25
.b
*-------.----..-,-,-m-m....--,,---.--,..%,--,,-,
w---y---.--
.,.,--..y3--.
---y,.,,-.---,,,---.-------w-~,.-..,-------w--
u e
TABl.E 1150. Activity in ret..e : nroups*
er, the very high setb. :y.* c-.:*a en water and the s:rong tendency of a:m::-h ric icdine to c! ate out Grouo Acuvdy en surface quickly redu:cs t..e amount of iodine in the air. Cesium. less velatile. is not expected to be s
i 2.97 x 10 Ci present in the air in a sign!!!: ant quantity. On the other hand, the solubi!!ty of xenon arid krypton is s
il 4.47 x 10 Ci very low; these gases wi!! be found almost entirely 7
111 4.6 x 10 Ci in met air.
To sunwnanze, neary complete release of noble s
IV 1.61 x 10 Ci gases, iodine, and cesium from damaged fuel is ex-V.
3.85 x 10 ci pected, even if the temperature is below the molting s
point. Significant re' eases of teGunum, rutherwum, s
VI 6.34 x 10 Cl and more refractory materials will occur only if the 98"I' 8
Vil 2.69 x 10 Cl the noble gases wiR be found in air, and most of the i
Vill 4.30 x 10 Ci other !!ssion products will be found in water.
s Total 5.11 x 10' Cl'*
- A few elements of low total activity, notably Fe. Cu.
As, and SD nave been arbitrarily locates a t the beasa of
- Analyses of sam;:les of conta.nment air, reactor j
meetine point.
- Totaa does not ouite scree with calculated total coolant water, and auxi!!ary building tank water are actnnty necause of roundine.
summartzed in Ref.197J Reactor coolant at:alyses show between 7% and 15% of the calculated Irwen-and actinides is given in Table B-56 for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after tory of 'edine and cesium isotopes to be in the r(.
cociant. If these measurements t.re corrected for
{ - }
shutdown.
- D A detailed r#=rmian of the fission product-dilution by water from the borated water storage release pathways begins in Section 5.8 of this re-tank, the fractions will be a factor of 3 higher.
port where a short summary is included. Radioac-Results for refractory materials show great variation.
,tive material released to the reactor coolant may A sample taken on April 10 was analyzed by four la-d have been partially flushed to the contamment boratories. There was a large variation frem fabora-through the open PORY (RC-R2). Some of the ma-tory to laboratory, indicating low confidence in the terial may have been flushed to the contenment pri-results. Analyses of krypton and xenon, isotopes in or to the contanment isoistion and then pumpedito
- the contanment atirn-disse also showed censider-i
~
the auxiEary buildng. However, the coolant may able variation. However, based on the most abun-t 83Xe), there seemed to be have contaned only a minute Itaction of the total dar.t isotopes (8%r and 3
activity at this time;it is highly I,ivieM,;. that a sie-09% to 62% of the core irwentory of nob!e gases in ruficent fracten of the coolant was released before the containment air. Only 2% to 3% of the, iodine the reactor buildng sump pumps were shutdown.
and cesium was found in the auxiliary building tanks.
There is an unsubstantiated possibility
- that mere On August 28,1979, a hole was dnlied into the water leaked to the auxiliary building after pump ' reactor building and samples of sump water were shutdown. This leakage would have termmated removed. Analyses of these samples showed 22%
when the reactor building was isolated after 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to 46% of the core inventory of iodine and cesium to i
be in the reactor building sump water.1se in addition 56 minutes, Most of the materlei thshoo' out of the RCS prob-to lodine and cesium, very smaD amounts of Ru, Zr, ably remained in the reactor.bdding. Some add-re, Sh, La, and Ag were found. As' expected. little Sr was found. At most, me amounts w...,,eiM-80 tional material may have volatfized from the makeup tank. Aside from thiese lossres, which are not ex-ed to a few millionths of the core inventory. About 4
pected to be very large, estinates of the total activi-0.02% of the core inventory of12s*Te was found.
As of these sample analyses were corrected for I
ty released from the fusi can be made by analyzing decay of the radionuclides to the time of analysis.
the reactor building air and water samples, the reac-tor coolant, and the auxiEary buiding tanks.
This correction procesa is certainly more accurate lodine is quite volatile, and it may be stoposed than the analyses memsefves; Le., the accuracy of mat a significant fraction is found in the air. Howev-the estimates does not depend on the accuracy of i
v.
1
,O i
the c;-
- u. r 'tfon. T.-h'a :: 5;* is a re-- * '.:.-
- r ; c!: :;
- ":! m m g::m:s:.
M
- :.ntion several as-ed by tt.s b:.~n", ':.:.*.
of thc rt :.:p c!'.c!ati!as.
~
pects cf the hydrc;cn 'ptct:am* are discusse::.
\\
The foaowing subjects cre; treated in this secticrt
, j Findings From thcce results, one can cautiously' conc!ude
- 1. hydtcgen production,
- 2. hydrogen accounting.
that between 40?. and 60% of the core inventcry of
- 3. calcufatien of bubb?a size.
re' ease secups I '!! was released to the coolant; that
- 4. removal of the bydrogan bubble, and only a small fraction of group IV was released; and
- 5. the hazard from the hydrogen bubble.
that only minute amounts of the remairung groups were released. The amount of refractory isotopes released is consistent with leachmg (see Appendix ' ' Hydrogen Production gy).
These data tend to conArm other analyses of Two possble sources of hydrogen are con-core damage. The data on, radioactivity released sidered:, metal-water reactions and radiolysis. Oth-are too sparse and variable for a precise conclusion er conceivable sources incfude oxidation of UO,3 to be made cri the amount of core damage; howev-wfuch has not been invest! gated. The productiort of er, the fc!!cwing conc!usions appear to be support-hydrogen frcm metal-water reactions is known to have been targe; therefere any hydregen frem other ed.
mechanisms is expected to be s.M h com# son.
. t About 50?. of the reactor core was damaged suf-
.Radiclysis is not expected to produce large fciandy to ra!aase the most velatife fission pro-amounts of hydtcgen. It is investigated because the
- ducts, possNty of oxygen produckn was censidered at
- 2. The low fract'ons o.f taffUrium, ruthenium, and the time of the accident. If oxygen had been strontium indicate that no significant quantity of released, the hydrcgen that was tracped in the
. fuel reached the melting point of UO (52007).
reactor coclant system ceufd have become famrn-3.The amount of refractory isotopes i the reactor able.
5 coolant is consisteInt with leaching, i
Metal-WaterReacdca
- d. Hydrogen Production, Removal, and Hazard Many metals are oxidized by watar. The react.ion
- is very slow at b temperaturas for mest meta ls.
Intrcduction Beth steel and zirconium are oxidized at an incress-
'ng rate as the temperature rises. The oxidation of One of the surprises of TM!-2 was the format!on i
of large amounts of hydrogen from the reaction of zircornum, the major constituent of the cladding, cc.
~.
TABLE 11-57. Total volatile Isctopos released from core n[
isotece (fraction of core inventory) 9
- Cs h.
3 Kr Released
'3'l (8MT,,)
'3'Cs
'33 To Xe
/
8 Environment 0.01'
)(
3 R8 Atmosphere 0.46 C.22' O.48' O.34' A
RS Water 0.14' O.12' 0.C8' X
RC Water 0.03 0.C3 0.02 Aux. Sidg. Tanks Totals 0.46 0.39 0.63 0.44
'See Ref. See ashes indicate few values igenerally tesa than 1*.)
- est estimate from data in Rei.19T.
Average of CDservations.
c
$27
~
o 4
Duzz Powrn COMPD'I Powra Bers.stwo 4a2 Socra Caraca Sruzzr. Cuan2.oriz. N. C. asa4a wwa= a.
a m. s n.
ses. *=osoevo.
May I.,
1981
.* ' j'.*3,',,'" 3 v
i.
tCb Mr. Harold R. Danton, Director g
Offica of Nuclear Reactor Regulation jd U.S. Nucisar Regulatory C & ssidu p
4 q bcS,
Washington, D.'C.
20555
\\
N t
,d,gO Attantion:
Es. E. Adansam, Chief 7
g A
Licensing Branch No. 4
.Ap. OY
.Q7 Re: McGuira Nuclear Station Docket Nos. 50-369, 50-370 Daar Mr. Danten:
Attached are can copias of McGuira procedura, A?/0/A/5500, "Is 1=ata of Failed Tual 3ased en I-131 Concentracien". Ihis is one of the i ple= anting pro-0 caduras for the McGuira Energency Plan and as such should be included with the other 1 planancing proceduras previously sub=icted en Tebruary 13, 1981.
3y copy of this latter, thras copias of this i=ple=enting procedura are being providad to NRC, Region II.
Very truly yours, Vf114 = 0. Parker, Jr.
GAC:pv Attachnant ec:
M. J. Graham (w/o accach.)
Mr. J. P. O'Reilly, Director (v/3 cys.)
Resident Inspector U.S. Nuciaar Kegulatory Co---4 ssion McGuira Nucitar Stati=n Region II 101 Marie:ca Straat, Suica 3100 Atlanta, Georgia 30303
/MS1WNIKy MXJ CCTL) 35c l sa snu a.+mw ro 5
W SdTWSS /
5 N II f G,%Ns,act.dlUc c.,sJu1)/Li3 l
A-F2,J meM uim c.c.4 yMJ' b / &c.49Aacs.
e a
..,.,./,j. ;.,.:<
..., j. ;.,. e fgg3. ?CW13. C*"Alf!" * ~' '~ s
' ~(1T D ' 30. i7 m / x/g
- e
, PRCCDURI ?'" IRA!!C3 Cha=ge(s ).
n
. ~ 7 :.'..
.... e.
.,a,,,-
..,.,,cezse..,,,.
70R 3 0. 0,E*_?' #I U 0
F.A-CN:
McGui.
P1cCZ:2U2I 5:U.Z:
Estimate of Tailed Tual 3ased on I-131 concan: ration P.!Z2ARD_.3T3.
N clae
- M S. M +1ea b DA"Z:
3f1a/0,f N
h 1 - 7/
w,de3::Wm Os/.A A
na: : -
Y 57/_ f/z7 L.:co-ciscip"- 7 1 vtav 37:
""M?C7as! A??2C7AI. (I? N2rf.ssia!)
(320) a.A 2:
A 2:
C.:
[
/d P/
8#
DA I:
N. 3!;
-/
/
m n r:.cas:
J cu, ) h......
3 /4 p/
j em cA z:
1=ri==.=d/ art-4 d.37.;
/
3.zviav;d/ App...ed 3r;.
CA zi h
e et d
(
e
_g
~~
7gg ggC01*d2.
3 "*.'
DM m 'C& DANT
w:.r.;. -. ~.. -
-s NUCLIM. SAIITT IVALUAC CN..CHICI LIST'
(
1-aMneipHETINI.N
, 1) 5 y CN:
b=
UNIT:
o73:3 rt,
t..: c. m ii - -
(2) W -?'L.ST A??LICA EI To:
A P / cl c I est--I
.3 (3) SA::..:. E7ALUATICH - PAIT A The 1 = :o which this avalua: Lou *s applicable reprasa=:s:
Tas No K A cha=ge to the s a:i== c; p =ceduras as desc=1 bed i= the 75A7 or a tas: or expa:1=a== =c described i= 6a ?SART If the ansva- :s the ab=va is "Tas", a::ach a de: ailed dese:1p 1== cf :he 1:e=
being avaluz:ad.a=d an ide=:ifica:i== of the aff ec:ed see-2:=(s) cf he FSAR.
(4) SA : ; UA*.UA!!CN - ?A2" 3-Tes Ne X
- 11 dis i:e= require a change :s :he 's:s:1 = Tech =1:21 Specifica.1==s?
.! the a=sva:
o da a*:cve is Tas," ide=:ify :he specifica:i==(s) aff ected a:d/or a:.ach :he applicable pages(s) vi h de cha=gs(s) *-di:a:ed.
(3) SA::..i E7ALTAT'"CN - PRT C As a resul: of da 1:a= :s which his evalua 1c= is applicable:
Tes No %
W'" de probah:7 of a= ac=1da== previously evalua ed i= the TSM be i==: eased?
Yes No X Will the c==sequs=cas of a= accide==. previ:esly evalca:ed i=.he ISAR be i=craasad?
Tas No X Ezy the possib:7 of a= ac:ida== vhich is diff ars==
-ha= a=7 already evalca:ad i he TSM be crea:ed?
Yes No X W
- he probab:7 of a =alf==:1c= cf aqui;=e==
i=per.a== =c saf e:7 previ=usly evalum:ed 1=
he 752 ha i=creasad?
Tas No 7" de c==sequescas of a =a.15=c:1c= cf equip =e==
i=per:a== :o saf a:7 previously eval =a:ed i: he FSAR be *- ~aasad?
.X.
May the possib:7 of alfu==:1c= cf equi e==
Tes No 1=serta== =c safa:7 diff are== :ha= any already evalua:sd i= the TSAR be crea:ad?
Yes No Y
W'
tha =argi= cf saf a:7 as def1=ed i= ce bases :s a=y Ta-W S Seccifica 1 = be reducad?
- s a=y of he p; ace ' g is "Tes", a= "- eviewed saf e:7 d
If :he ansve:
ques:1c= is i=volved...fus 7 :he c==clusi:= ca: a= u=:eviewed saf a:7
~
ques:1c= is or is =c: i=volved.
A::a d ad,di:1::a1 pages as =acessa:7
(
(5) 72I?pm 3T: Mid~.d X N.A.A DA=:
4Ia/ei
'M J-m,+
- 2f., /A=:
Al!C: 97 (7) 12...IWD 3Y:
/
V (3) Page 1 of I
A?!0/A/f300/33 DUII ?v u. CCMPANT McGUIII NUCLIAR ST.ATION ISM 07 7A__ID IUIL 3ASID ON I-121 CONCIT"KAI CE. '..
~
1.0 Se==s 1.1
? "?-43 raac:or coola=: =adia:ic: =c=1:c: has ala=ad.
1.2 IIET-18 mac:c coola== fil::: 1A radd s-d-=
=c=1:== has ala=ad.
1.3 1 "7-19 reactor c=ola=: fil::: 13 radia:1== =c=1== has ala=ad.
1.4 A=y plant ec=di:1m= i= vhich the apara:s; vould suspac: failed fual'or va==
a=' as d=a:a of ha a==u== of failed fuel.
2.0
~~adiat a Acti=
2.1 Au:c=a:1c No=a 2.2 2.2.1 Cb.21= a ch d *:- ; s =pla =f ha ::::::: c:cla=: i= c:da: ::
da:a=i=a :he :-121 c =ca==a:1:= cf.he c=cla=:.
2. 2..".
C=ca ha I-121 c =ca= =a:1== is 2:=... f : :he raam::: ce ia=
da:a.:- d a which of the fc11cv1=g fou casas bas: describes 1
.he p;as=== f:el c=-dd -1==s.
NOTE:
A.
~ha -
- ars =h-d-*d by ust=g.his p:: cad =a ara a:
bas =, as:d=a:.as c=17 3.
All fc- ' = = que:ad a:a based upc= aq"d' di:1= f
pcus cars indd*a.
If fuel da=ags is suspa=:ad := Ezva accu =ad d=1=g :1=as of reduced pova: c
=sa: ' ha d=a = f
.~
sig d'd-*-: pcva e's ge, the em:s ind1=a i=ve==:7
=us: be c=pe=sa:ad ac:::d' gly by usi=; I==1 s=a 4.2.
- is is :ha c==ac:ic: fac::: T.
C.
All va.luas giva= ara
=r-*' d:ad = vel =as of==cla=:
a: =m=al reac=
=c'.a=: systa= ;:sssura a=d :s=parz-tu=a.
To em=s== f== a:har NC syste= :e= para:=ss ::
reducad NC sa=ple :.._ para==as, esa I:=1osura 4.1.
- d*
is ha c =ac.dc= fac:c: Z.
13~3 D.
The decay of T
.m :-121 '..as bea= taglac:ad as a
d - *ig if d -*": i= 0' d
- a=alysis.
/
(
2 r-E.
Iodina sp Vd g =ay oc = af:a a shu:dov: c:
d
(
sig=ift a=: povar cha=ga.
Oa:a f := ocha Westinghouse p12=:s has shown tha: :he ied1=a sydVd g process has ban: observed 'to oc:= duri:s a parted of 1 :n 3 days af: : che cha=ga or shu:-
dov=.
Eovaver, che spika san =s :s peak d=1=g the parted f:ca 4 to 8 be=s af a
- ha.cha=ge.
I-131 et=ca==:a-J.ons ca= inc:aasa by a fac:m: of 2 :s 25 above :ha aq"d' dh:1= lavals durt=g hasa :i=es, al. hough a=
d--
aasa over a fac: : of 10 is== usual a=d would o=17'ba sea: a: a shu:dev=.
I:::aases by a fac:n of 2 :n 3 ars. gi:21 f== a si.=1fi---:
povar dae:aasa (i.e.,1002 :o 3C: p= var).
Do :::
d=i=:arpra: :his :a=; crazy cha=ga fc fuel f211= a
- f :hara is :o c:har evids=ca of fuel da= age. C:har avida==a cf fuel da=aga --- be : _s:i=:ad by a=y d-dd -*:1 = of i=adaeus:a c :a c=c' d 3,1:esa parts d-dica:i==, high i=cora tha:=oc:uple i= dita.1:=. a::.
/
T.
If esti=a:as for fusi fai1= a a:s =sedad f:: fuel w dd=1 =s acha= :ha:. hose c=varad by -h f=ur casas described balov, or if=== ac= 2:a fuel fa11== da:a is =sadad, saa Sa=i:1 = *. 2.7 of his procedura.
G.
~he fs11=v1=g four cases c ver a very b;:ad :s=ga ed ec:a c=
dd=1==s.
Choosa :ha c:a.ha: bas: sui:s :he azisti=g c==d1:1c=s.
E.
Cha=1s::7 sa=ples should be :ake= as soc = as da= age is suspec:ad.
2.".2 Casa I - No::a1 Opera:1==
2.1.3.1 he===di:1c=s which par _ai: := Casa I - ::c:=al ope:atics are as folievs:
2,.2.3*1.1 No =al raze::: cpara im: a: a=y pcva:
c shn:d=v: vi h _c cusual c==di:1:=s pris: :s sh::dev=.
Adaqua:a ::: c:cli=g has b an: - d -- = ' ad.
i
-~
3 3.2.3.2 If the abeva bas: dese 1 bas the cc:a ec=di:1c=s, use the f=11cvi=g fe:=clas :s cale-da:a the rs=ga of failed fuel values.
Ira.lus:a
==ac:1c= fae:n s I a=d Y by usi=g I=closuras 4.1 a=d'4.2.
2.2.3.2.1 -(Wasured I-131 c==a==a:1== uC:/=1
-3
+ 3.5 x 10 u CI/=1)
I T
= Nu=bar of,failad pf=s (hz. expected a=d bas: as 1= ace) 2.2.3.2.1 0'.aasured I-131 ce=ca==a:'== u C /=1
+ 4.9 x 10~3 u C /=1)
I T
= Nu=har of failed pi=s 03.1=. expec:ad) 2.2.3.2.3 (basurad I-131 ce=ce==a:1c= uC /=1
+ 1.3 :C /=1)
IY
= ?a :a=: fallad f a1 (v.a=.
a=pe==ad and bas: esti=a:a) 2.2.3.2.4 (basur d
- c==:a==a:1:= a C / 1
+ 2.3 u c./=1)
IT O'l
-- =_?ar:a==_fallad. fuel.(P.1=. axpec:a'd)
V NOTI:
Typ'- valuas is: I-131 c== a==a.i== '.= u C./=1 is:
-1 a =c = ally opera d g pla== are be:vas: 1.0 x 10 a=d
- 4.0 x 10-2 uc./=1.' Sasa vduas a:s basad o=
ha :
3ze:::
coola== I-121 a :1ri:das a= paria==sd by ha Cic= a d !:o::.:
Pla=:s.
2.2.4 Casa II - h ercscapi: Clad Da= age 2.2.4.1
- ht es=di.1==s which par:=d- :: Casa :.-
l'acroscopic clad da=aga are as f=11=vs:
2.2.4.1.1 Nor=al raze.:: cya:a is: a: a y pcvar, c: shu:d=v: vhara sc=a =a
'-=-d-=* clad fall = a (1.a., a *,cesa ;a= _= 1:::
i=disa:1:=) er a f1:v ' du=ad failura is suspes:ad.
The c=:s has adaqua:a c=olf:s and =o sig=1f1=a== fu,a1 =verta= para:ura is obsarrad.
I h
g
s.-
r.
- 2. 2.1. 2' If hs ciovs bac: dese:1bes 63 c= a essdi:1 ass, use tha' fallowi=g is:- 'as to calcula:a tha :ange of failed fuel values.
Ivalua:a es : action fac:::s I a=d T by us1=g hel=surs 4.1 a=d 4.2.
2.2.4.2.1 (Maasured I-U1 c=nce=:: :iss uC/=1
+ 5.5 x 10-2,g;j,y),g,7 Nusher of failed piss (Max. ezpac:ad)
=
2.2.4.2.2 (Maasured :-131 cance==:stice u C/a1 16.5 x 10~I u CI/ml). Z T
+
.% saber of fallad p1=s (3as: es.ima:a)
=
2.2.4.2.3 (Maasured I-131 c==cas::a:i=s 4 C/ 1
+ 27.4 x 10' uCI/ml)
Z. T Number of failst p1=s (Mi=. arpactad)
=
2.2.4.2.1 (Maasured :-131 ::sca==:a:1== 2 C/=1
+ 27.9 :C/=1) *I+!
Per:s== failed fual (Max. e=pec:ad)
=
2.2.4.2.3 (Measured I-131 c ses== a:iza u C/=1
+ 83.7 u C/31).I T
'h 7arsa== fallad fuel (3ast as.1=a:a)
=
v 2.2.4.2.6 (Maasu ad. -131 c:ncatt:s: ion aC/ml
+ 139.5 h C/=1) 1Y
?sreen: failed fuel (Mi=. azpec:ad)
=
2.2.3 Casa
?--
- Savara Tual overtempers:u a 2.2.3.L The as= W ans wh1=h parta1= :s Casa I Severs Tual Over a= para:urs a s as f=1;=vs:
2.2.3.1.1 2C: :ype accida=: *.. ara thera has ben:
an ahns==al shu dov= a=d i: is susyac:ad
-ha: the fuel has been a: lass: pa==1 ally.
uncaverad f : a ; art:i :f ::.=a ;;sa: : -ha=
a few =1=u:ss.
Ted" g i= ha :::s is da: acted by high ines:s :ha==se=uple :sadi=gs and loas af =a.3 = :s sa:ura:isa.
Tual 1
o clad azida:iss is da:ac:ad by azzass h7 relas in da c n=a1==an: or 1= the d
- sactar escla=
sa=ple; however,== fuel
(
.a m =, o s.,ac=ad.
e i
5-2.2.3.2 If the above hast dese:1has the esta asuditions, I
use the !ollowi=g ia=ulas to cal =la:a tha :a=3a of failed fuel values.
Ivalua a cor ee:1os fae: ors Z and T by using Z= closures 4.1 and 4.2.
~
2.2.3.2.1 (Maasured I-U1 es= centra 1mn u c:/=1
+ 2.4 u C/al)
- I Y
= Number of failed p1=s (Maz arpested) 2.2.3.2.2 (Maasu' red I-U1 concan::a iss uC/=1
+ 2.9 u CI/al)
- I. Y
= Number of failed piss (3as: estiasta) 2.2.3.2.3 (Maasu.ad I-U1 enneantration uC/ 1
+ 3.2 u C/ml)
Z T
= Number of failed p1=s (Min. expe=:ad) 2.2.3.2./. (Maasured :-01 esssa=::a:1sn C/ 1
+ 1233 uCI/ 1)
= ?ar=a== failed fuel (Max. expec:ad) 2.2.3.2.5 ('.d.aasured I-U1.c==ce== s:1on u C/ '
O
+ '.333 u C/=1)
ZT
= 7arsen: fallad fuel (3as: asst.=a:o) 2.2.3.2.6 (Maasured I-131 esacan::a:1an u CI/mi
+ 1673 uC/ni)
Z
- T s
= 7er:sa failed fuel (. szpas ad) 2.2.6 case IT - Fual Mal:1sg 2.2.6.1 Tha essdi:1sas wtd.ch parta1= =a Casa IV - Tual Mal 1=g, are as isllows:
2.2.6.1.1 Severa as=1de== vhara.hara has been a=
abas mal sh=:dow= and the es s is.u =.-vered for a lots period af ti=a.
!===
s : hor-measuple :ampara ::= :==dd s ars abava 1300*7 far a 1=ng parind of 1sa.
Tual sal.iag is suspected (i.e., fuel :empera-zurs essands 5000*?) a=d is verified by the i=ah111:7.s operata the t=ests i=s::u-senta:1sa sysuas p;sya:17 l
-e-
' t.
2.2.6.2 If :ha above bas: descri'ses :he em:a coe 'd:Lo:s,
(
usa che fc11cv1=g f==ulas to cale=12:a :he fallad fuel values.
Ivalua:a cense:i== fac: ors I a d Y by us1=g b elosuras I,. 1 and 4.2.
2.2.6.2.1 (basured I-131 cc:.cas::::1 = u C2/ 1
+ 3.! u CI/=1)
I. T
= Nu=bar of failed piss (3as: esti=ata) 2.2.6.2.2 O.'. assured I-131 cc=casc 2.1== u C2/=1
+ 2790 uC2/ 1). I T
= P an a.,: of fallad fuel (3as: es:t=a:a) 2.2.7 If fuel c==di:imus o he: tha:.hosa described a': ova e=is:, c:
if a =cra detailed f allad fusi es:1=a:1c= is dastrad for at ha: a:arga cy or =c=21 cpara:1 :, cc::ac: :he app;=pria:a
- Jas:isshousa yacpla balov 1: -ha orda: listad us 11 es: ac:
is =ada.
,2.2.7.1 E=argn=7 7' *-: C:=di:1==s - hergn==y Aaspo:sa 2aa= *Jes:1=ghouse, ?t::sbu -A, Pa_=sylva:La 2.2.7.1.1 Diras::::
Ea=k Ruppa; 412/0!6-3611 *Jeri q
i
-f 412/366-6751 Ec=
y 2,.2.7.1.1 Depu:7 Dirac:me:
Ims Lehr 412/236-3401
- Jerk 2.1.7.1.3 Te *-
d-*' 'Suppon P.a=aga::
".s A=da: son 412/373-5766 7c k; 412/327-3239 E==s 2.2.7.1.4.Y.a:ard.als assig=:
Wally Chubb 412/373-4364 **c:
2.2.7.2 No=al Plan: Cc=di:1==s 2.2.7.2.1 Sou.ha= 2agissal P.anagar - 5:ava *::gd:= -
404/883-3900,
- c k 2.2.7.2.2 Vas:1:ghouse - auka lagrasasta:1va - F.ika
- " a: - 412/373-3160, 7c k 2.2.7.2.3 P.a:::tals 2 asis:
- Jally Clubb - 4*.2/373-4354,
- ork
3.0 Subsecuan
Ac:Loes 3.1 Tollow up as =ecassary tri:h *Jas*' ghouse - ?i::sburgh, ?s=s71va=1a o
depe=d1:3 on the pla== si: a.1cs.
(
-7 4.0 Inclosures 4.1 Densi:7 Cs::actie Tac:c, I, for :tc ;a=pa:a:ura Changes 4.2 Iodise 131 !=ves:: 7 Cs :sc:1c Factor, T, !== reducsd pova opera:ic c: fs t1=as of power cha=ge 4.3 Ixa=plas O
e 9
9 9
9 8
9 9
9 e
0 5
O ee e
=
9 6
0 9
9 e
e e
9 e
p 9
e G
4 e
em 9
e O
O e
' r.
A7/0/A25300/33 f
'* * ' a sur a 4.1 Det.si:y carrac:1== Tac:=r, I, fs; NO Ts para:ure Cha= gas 71=d :he appropriata N0 System campara:urs at :he time of accida=t.
Ti=d the approxi=ata camparature at which the NC samplas are takan. The in arsaction of both nazabers is the density cor; action fastar, I.
NOTZ:
No-/.a1 NO System sampla em:pera:::a is app;szi=ataly 90*7.
Use this tamparatura 1.f no oths; i=fs mation,is available.
NCS Sa=ple Ta=peratura 'T l
80 90 l
100 t
100
.996
.998 1
- }
130
.983 *
.983
.987 200
.966
.968
.970 s
150
.945
.947
.549 w
300
.911
.9:3
.924 330
.894
.895
.597 sa
.I3 400
.462
.864
.363 l
e 1:
450
.st7
.8:5
.a20 i
" 1E l
500
.757
.715
.790 s
- 3. _1, 550
.73i '
.740
.741 d"
360
.713
.729
.7:1 a
I*
370
.717
.718
. 719 5
580
.706
.708
.708 u
590
.693
'.694
.595 600
.680
.681
.683 e
k
e Pago 1 of 1 AP/0/A/5500/33 helosu a 4.2 Iodi=a 131 !=ves:ory Cc=ac:1==, T, fc Reduced Pava Opera:1c= c f== T1=es of ?:ver Cha:ga 51:c.a:1 = 1 70 caract.for coro Iod'"a I=ve=:ory if fuel da= age is suspected to hava c:c==ad durd=g H es of a=y pova: level excep: 0: whars :hn pcvar laval has no: cha= gad graa:a
.ha: -110: vi-Sd - the las: 22 days, usa :ha fc11=v1=g aqua:Lon.
100 Y=
Full Power a: :1=a of faile.ra where T is.hs c==ac:1 = fac:s; to ha used i= See:1:= 1.0.
?-
ple:
- ha p' has *ses= a: 35. full pcvar f
- ha las: 30 hys whe= fuel damage 1.s suspected.
Theradora:
.~ i 100 T=
= 2.86
~5 O.
Cl.d:i== 2 To ec::ac: fc; cars iodi=a if fuel da= age is suspec:ad. :s have cccu=ad a:
- 1=as o c a: -ha= fi: 51:ua.1== 1 above, usa.:hs f=11:vi=g aqua:1==.
100
,= --
cid pcua: leral 1= *. (a' I ) + =av p ver laval 1= : (1-s
- )
stara:
T = cc=ac ics facts; :n be used i= Sac.i== 2.0
- 1d pova: level 1= : = da : f " pcvar bef=ra :he pova: cha=ga l
cas pova: level i=, ; - :ha : f; t ' pcvar af:a :ha pcvar v-' ga a: which I
.1=a.he fuel fa11= a has occu=ad 4
A; = is :he decay c==sta== f== I,y which equals.CS64 day
- 4 is :he =ad12= sina :o -.aka a pcva c'ange plus tha i=a af:a
- ha pcvar changa
- 11 dz= age is suspec:ad.s have oce==:ad, i= days.
Iza=pla:
If 1: cek 1 h=urs :s -.aka a peva,: chassa a:d h= age was suspec:ad i
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> af:a
- ha pesar cha=ge.
9
- - -- + 10 = 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> 1
Pageb,cf8 A?/0/A/5500/33 p
- tosura 4.3 Iza=plas,
?: ble= 1 a.
Power level has haa= dae:aased f = 852 :o 502.
' b.
This power cha=ga scok four hours a=d oc:uried be:vas: 1200 a=d 1600.
8 T
a: 50% is 570 7.
AVG c.
A: 1800'a loose par:===1::: ala== goes off daddsa:1=g a loose ohjas:
i=.he as:a.
The ranc:c is not tripped.
d.
A Chs=is::y taa= is i=nadia:aly dispa::had.s :aka a sa=pla NO Sys:e=
~
as failed fuel is suspec:ad.
a.
Cha_is: 7 sa=pla d-Ad -=:as I-121 c=====::::1:= is 10.0 'u C./ 1.
Par: 1.
Da:a==i=a he bas: esti=a:a of the =u=her.of fai*ad p1=s.
?ar: 1.
Date: d-a :ha bas asti= ate of pa----
- ailad fuel.
felu:1cd 1s Caia II, 5:ap 2.2.4
- d=
Use equatio= 1.2.4.2.2 !== ?ar: 1 Usa squa:Lo= 2.2.4.2.5 fs; ?ar: 3 t bg en-- a=-
a '---
/ t
-aasured t" w
d Par-1.
. I T = Nu=har of failed pi=s
(
16.5 x 10 u C./=1 Da: : d a I:
Z::losura 1.1 T
is 570*7 a: 5C.
AVG Assu=a NC3 Ea==la "'a= para:ura is ~0*?
harafara, I =
.7'_3 Da:ar d a T:
Taa'osura 4.2 1
A.'=.0864 day -
s.
- =
(4) - (2) = 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> T
'/
3.a=s=bar, : is :he =adia=.1=a :o _aka a power chz=ga plus :ha differs==a ha:vas:
\\.
- he :i=a wha: :he da= age is suspec:ad a=d tha' :i=a tha =av pcvar laval is tached.
heursxk..,daZ.s.,,.57 days C = var: : es days
- = /*
a O
4 w
-,s e
1
.-n
Page 2 of 8 i
AP/0/A/3500/33 I:closurs 4.3 Exa.plas 100
,, r,
- c.08 4 u, 33 c.m u,37
_g
,3.
.,c.Out a7 3 c.m u73 e
'00 T=
= 1.183
~
83(.9857) + 50 (.0143) 1I.
10 u CI/~'
(.718) (1.183) = 51.3 =52 fzils:1 pi=s Ex:: 1.
._sae:
16.5 z~10~2u CI/=1 (O1
.Fmas= s:! I-131 C===s==-z:1 = u CI/ 1 I
,.=
, ails:. u.a1
- car:
4.
83.7 uC;/zL 10 uCI/ '
Sa... e a C;/ _1 (.718) (1.183)
= 0.1% fallasi fuel A:sse:
o e.
/"
(
e e
Pago 3 of 8 f
A?/0/A/5300/23 I= closure 4.3 Iza=ples
's able: 2 a.
De reactor has jus: ::1pped i= seas:17 fro = 100 power due to a malfu=ctioning iss=t=na==.
Thara vers =a u=. usual es d'.io=s pris to tha ::1p.
b.
T is =ow 537*7 a: 0% power.,
gg c.
3 a opera:c;,
fa hav1=3 :o rasso= to suspec: failed fual, d.s ee.:1ous abou: cha a===== of failed fuel presa== av f=1*:vi=g ths
- 1p.
d.
A Che=is=-r :sa= is san: m aka a= NC sa=pla 10 h=urs afta
- he =17
'"ha Cha=is::7 sa=pla 3 ras a= b-121====s===a:i== =f 0.0 x *,0 ~.:C /=1.
w 1
a.
(A :7pi-*' value f== c =c-_ ally opera.1=g pla==.
See Nota -da:
lQ Case I, 5:ap 2.2.3.2.)
L'l Che=is:ry persc==al also
'"dd -*:a cha: 'NC-sa=ple :a==ars=== is 100 1.
f.
expec:ad =.=nbar of failad faal piss.
Par: 1.
De:a-- e* :ha -=~'
Par: 7.2.
Da:=
' s.he =a=i==a expectad par:a== failed fuel i= :he c=ra.
. Sciu:1==
~.
"his is Casa I, 5:r.p 2.2.3 Usa squa:1:= 2.2.3.2.1 f== ?ar: 1 Use equa-h=.1.2.3.1.3 for Part 2 I
Maasured I-131 e======s:1== uC*/=1 S,
.,,,,,,,,, a, s.s.
., z._.,
a,,
e 3.3 x 10 uCI/=1 d
De:ar- '-a I.
I==1= sura 4.1 3C Ta=pers=== is 537*7 a: 0 SC sa=pla :-rars==a is 100*7
. Deraf=ra, I =.732 l
-r,--
r.~,-
,---a..,
,e,,,,-,,
.,,,,,,,,ng,,,
a w y,.
Pago 4 of 8 AP/0/A/5500/33 s
(
I closura 4.3 Izz_plas Da:ar-4 a T:
I==1csurs 4.2 S1:=a:ic= 2:
e = 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> x_1 dav =.5 days 44 hrs.
~
100 T=
100 -l(.0364)(.5) + 0
- f(.0864) (.5)l.-li 1-e e
w
.c
! = 1.044
, NCE:
I_2 : = 0 c: a sa=pla tas :2ka= d edla:aly, T = 1.0.
v ar:,.
C/=1 (..,. 2) (
0u.) = 4.4 2.0 = 10 2
4 3. 5 x :.0 a C/=1 c
4 :s 5 f ailed pt=s answer Matsured I-131 C= ce== 2:1== u C/=1, y. T
failad fuel 7
1.3 u C/=1 2.0 = 10' u C/d
(.722)(1.044) =.0085 : failed f a1 1.8 u C/ 1 4
"'ha above -
- a:s a:a d -d d -= :1ve o f =c:=al cp arati==.
A:s.ra SC I:
I-121 sp9'*g =27 be a p hie = hara.
San Scap 2.2.2, Sc:a I.
$ngo5of8 A?/0/A/5500/23
- v. 'asura 4.3 Iza=plas
?-chis.: 3 a.
Favar level $as bes: be:ves: 502 a=d 652 fo: ha las: 30 days and is prese=:17 a: 602 a: 1800.
~
b.
T is =575*7 a: 6'0: Power.
g c.
I: is desi=ed :s ses if a=y sig=1 fica== failed faal exis:s i= :ha c= s even though =m ab=a =al oc=u===cas have.aka= placa.
- d.
A: 2200 the same day, a che=1s: 7 sa=pla is :aken =f the 3C sys:s=.
.n The che=is: 7 sa=ple '-ed -=:ss I-121 c==ca==:a:i== 's 3.9 x 10
- uC /=1.
e.
?ar: 1.
De:ar d a :he bas: es:1=a a of :he==ie
=f failed pi=z.
Par: 2.
De:a: ' e he best as:1=a:a of :ha : fa11ad fuel.
- x
~
i:clu.1==
- is is casa I. Scap 2.2.3 Usa equa de: 2.2.3.2.1 fs; Par: 1 Use equa:1o= 2.2.3.2.3 for Ps : 2
~
. Measured I-131 C==es==s:1== uC /=1
,y
,,,.... a_ w..,
,c a
.. a-a..
.,a e
3.5 x 10' uCI/=1 De:ar dae I:
I=closu== 4.1
- AVG is 375"? a
- 6C: peva:
Assura NC3 sa==la :s=p. =f 90~7 "haraf=ra I a.7*.3 Date-das T:
I=cissu== 4.2 Si:ua:Lau 1
~
T = 100 = 1.67 60 3
Paga 6 of 8 9
6 A7/0/A/5500/33 E=closura 4.3 Izamples
-2 3*9 x
~'O u C'* /~'
Par: 1.
(.7U) (1.67) = U.27
,3 3.5 x 10 uCI/mi
=14 "miled p1=s Answer
~
i
\\
f Ch Esasured I-131 Canese: a:1 = u C~/=1 I,,.,.,ai.ad.,.,a1
,a..
1.8 uC /=1 3.
x 10 uC/d
(.713) (1.67) =.325: 'a11ad fuel A=sve-
~
1.8 uC/mi
~
Se aheve n=mbers a=a ac=aptabla f== a =c =sily opera:ist pla==.
e e
e e
y e
]
6
Paga 7 of 8 A?/0/A/5300/33 I= closure 4.3 Iza plas 7 c' ele = 4 a.
S a u=1: has bea= a: 97 povar for a=== h whe a depressu 1:a:1==
of ha NC sys:a=l occurs.
b.
Da reac_c ::1ps..
c.
Etavy vibra:i== is obsa:vad i= :ha NC pt. ps.
d.
Da==ccouple ta= para:uras eva 1000*7 are i=diez:=> '- -ha cera.
\\
a.
1.EG 48 and IE C 13 have gens off.
f.
Safa:7 !=ja:.1c= vas delayed a=4 1 is suspe::ad :ha :cra was====vered be waa: 30 a=d 60 _1=u:ss befera sufficie== reac::: vassal wa:a: level was regai=ad.
g.
The i====e iss::.=a=:a:1== sys:a= is still epa:sbla.
- 1 h.
The NC sa=pla d-d'-':as a I-131 c==:a=: a:1 = of 33Q0 :.: C /=1.
1.
A Che=1s::7 sa=pla is takan d-adia:aly (vi:' db the hour) af:a
- ha
- 17
?ar: 1.
Ds:a '-= chu -*
d--
a=pae:ad =+-har of faile.d
'-e.
~
Far: 1.
Da:::- '-a tha -'*-
arpas:ad of f ailed fuel.
Sc12:1==
S' ' is Casa '-*, Scap 2.2.5 Usa squa:i== 2.2.5.1.1 for ?ar: 1 Usa equa.1== 1.2.3.1.4 for Par: 1 Date: d a I: I==1csurs 4.1 NC Ta=p. T a: 0 pcvar b 53 y7g sa=ple :e= para==== c f 90*7 Assu=a e
S araders, I =.730 ~
".)ana
' a T:
'00 T=9
= 1.03 e
.-., e.
Pago'8 of 5 I
AP/0/A/5500/33 Inclosu a 4.3 Izamples 3800 uCI/=1 Part 1.
(.730)(1.03) = 1190.5 1.4 uCI/a1
=1191 number failed pi=s, =az. expe=:ad A=sw q
- a..
_.a 38bouC'/s1
?ar: 1.
(.730)(1.03) = 2.23. f ailed fuel, =a=.
expe :ed Assver 1:35 uC* Jai N
D m.
O e
9
/
g UNITED STATES
/
NUCLEAR REGULATORY COMMISSION o
{
WAS'HINGTON, D. C. 20555 s*****/
January 5, 1984 Docket No. 50-424 MEMORANDUM FOR:
8Elinor Adensam, Chief Licensing Branch #4 Division of Licensing l
FROM:
Karl Kniel, Chief Generic Issues Branch Division of Safety Technology
SUBJECT:
INFORMATION REQUEST - V0GTLE, UNITS 1 AND 2 Plant Name: Vogtle, Units 1 and 2 Docket Number: 50-424 Licensing Stage: CL Licensing Branch and Project Manager:
LB#4, M. Miller DST Branch Involved: Generic Issues Branch Description of Review: Unresolved Safety Issues Review Status:
Incomplete The Generic Issues Branch requests information from each applicant regarding 7) the status of Unresolved Safety Issues pertaining to that facility. This infomation is required to supplement the staff's discussion of these issues in each SER. The information from the applicant should include a summary description of the relevant programs and the interim measures they have taken pending resolution of the issues.
Enclosed is an information request that we ask that you send to the applicant within seven days of your receipt of this memorandum.
The applicant should respond to this request in a timely manner, so that we may incorporate this infomation in our input to the Vogtle, Units 1 and 2 SER.
Karl Kniel, Chief Generic Issues Branch Division of Safety Technology
Enclosure:
Information Request cc: w/ enclosures F. Schroeder T. Novak N. Anderson
(
P. Norfan l
N. Su M. Miller R. Silver i7
Contact:
T. Su, GIB/ DST 492-7477 w.,m.&
97 W
~
~
~
o Enclosure REQUEST FOR INFORfiATION The Atomic Safety and Licensing Board in ALAB-444 determined that the Safety Evaluation Report for each plant should contain an assessment of each significant unresolved generic safety question.
It is the staff's view that the generic issues identified as " Unresolved Safety Issues" (NUREG-0606) are the substantive safety issues referred to by the Appeal Board. Accordingly, we are requesting that you provide your justification for permitting plant operation pending resolution of these issues. This should include a description of any interim measures in terms of design or operating procedures or investigative programs that are being pursued to address these concerns. The justification should provide an overall summary of your position on each issue rather than a reference to various sections of the SER where related information is presented.
There are currently a total of 27 Unresolved Safety Issues.
We do not require information from you at this time for a number of the issues since a number of the issues do not apply to your type of reactor or because a generic resolution has been issued.
Issues which have been resolved have been or are being incorporated into the NRC licensing guidance and are s
addressed as a part of the normal review process.
However, we do request the information noted above for each of the issues listed below:
1.
Water Hammer (A-1) 2.
Steam Generator Tube Integrity (A-3) 3.
Systems Interaction (A-17) 4.
Seismic Design Criteria (A-40) 5.
ContainmentEmergencySumpPerformance(A-43) 6.
Station Blackout (A-44) 7.
Shutdown Decay Heat Removal Requirements (A-45) 8.
Safety Implications of Control Systems (A-47) 9.
Pressurized Thermal Shock (A-49)
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UNITED STATES
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g6091983 Docket No. 50-424 MEMORANDU'M FOR: Elinor G. Adensam, Chief Licensing Branch #4, DL FROM:
Mn. H. Regan, Jr., Chief Site Analysis Branch, DE
SUBJECT:
REQUEST FOR ADDITONAL INFORMATION ON V0GTLE Please request responses to the enclosed questions from the Georgia Power Co. so that we may continue our review of the Vogtle Electric Generating Plant FSAR.
These questions were prepared by Al Brauner of my staff.
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. H. Regan, (t fef Site Analysis B anchN Division of Engineering
Enclosure:
As stated cc:
M. Miller e5 9
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REQUEST FOR ADDITIONAL INFORt1ATION ON YOGTLE 311.4 Provide infomation pertaining to the proposed simulator, specifically:
(2.1.2 )
a.
where is it located with respect to the plant?
b.
what is the size of the staff assigned to it?
c.
how many trainees are expected to be involved?
d.
what arrangements have been made to control the activities of personnel participating in the simulator program since it is apparently within the exclusion / site boundary?
311.5 Provide the following infomation (include the distances and (2.1.3) directions in relation to the plant)
(a) Location of closest residence as well as type of residence t
(i.e., permanent or temporary occupancy),
(b) A listing of all communities within 30 miles of the plant that have populations greater than 1000 persons, and (c) Identification of the largest city within 50 miles of the plant.
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311.6 River Road, although relocated, still appears to run inside the
( 2.1.2 )
exclusion / site boundary according to Figure 1.1-1.
- a. If any drawings (figures) are in error, please revise.
- b. If River Road is located within the property line, provide infomation on the arrangements that have been made to control the activi.ty on this road, particularly during an emernency situation.
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311.7 Some of the numbers in the various population tables do not agree
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(2.1.3 )
(e.g., when comparing FSAR Tables 2.1.3-1 thru 2.1.3-16 with 2.1.3-17 and 2.1.3-18, the respective year population totals do not agree).
Please check for discrepancies and revise the t' ables in all documents for consistency.
311.8 No mention is made of the Ebenezer Church located within the LPZ.
( 2.1.3 )
Provide inforination on the size of the congregation, and whether it is active or not.
If active, indicate the frequency of services and/or meetings that take place.
311.9 Provide a listing of all prisons, hcipitals, nursing / convalescent (2.1.3 )
homes', day care centers, schools, churches, cemeteries or similar
~
institutions located within ten miles of the site.
Specify their locations (distance and direction), the number of persons ordinarily employed or in attendance (occupants, staff, students, employees, guests, visitors, etc.), and the capacity of each facility.
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UNITED STATES d-g 8
NUCLEAR REGULATORY cot.iMISSION
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WASHING TON, D. C. 20555
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S o't P NOV 3 0 1983 fi W f Docket Nos:
50-424 50-425 Q
MEMORANDUM FOR:
Eleanor Adensam, Chief Licensing Branch No. 4, DL FROM:
Frank J. Congel, Chief Radiological Assessment Branch, DSI
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 PLANT NAME: Vogtle Electric Generating Plant - Units 1 and 2 LICENSING STAGE: OL DOCKET NUMBERS: 50-424/425 RESPONSIBLE BRANCH: LB #4; M. Miller, LPM DESCRIPTION OF RESPONSE:
Review Questions REVIEW STATUS: Continuing The Radiation Protection Section of the Radiological Assessment Branch has C-completed its review of Chapter 12 and other pertinent sections of the Vogtle Electric Generating Plant Units,1 and 2 FSAR.
A request for additional infonnation is enclosed.
We are willing to discuss these items with the applicant if required.
We plan to make a site visit to resolve the questions in the attachment and will be contacting the project manager to arrange the details.
This review was performed by M. A. Lamastra, RPS/RAB.
Frank J. Congel, Chief Radiological Assessment Branch Division of Systems Integration
Attachment:
As stated cc:
R. Mattson D. Muller M. Miller 0 Lynch M.,Lamastra
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V0GTLE ELECTRIC GENRATING PLANT UNITS 1 AND 2 DOCKET NOS.
50-424 and 425 REQUEST FOR ADDITIONAL INFORMATION 471.02 As specified in Regulatory Guide 1.70, Section 12.5.2, submit a (12.5) detailed diagram of the station's health physics facilities.
471.03 As specified in Regulatory Guide 1.70, section 12.5.2, you should (12.5) mdify Tables 12.5.2.1 through 12.5.2.4 to include the minimum number of each type of health physics instruments.
In addition, provide information on the quantity and types of respirators pro-vided for the Vogtle Electric Generating Plant.
471.04 Figure 12.5.1-1 should be modified to include the minimum (13.1 )
staffing levels for one unit and two unit operation (see section 12.5, Regulatory Guide 1.70).
471.05 In accordance with the criteria contained in NUREG-0731, it is our (13.1 )
position that your organization chain should contain a qualified individual to provide backup in the event of the absence of the Radiation Protection Manager (RPM). The December 1979 revision of ANSI 3.1 specifies that individuals temporarily filling the
' RPM position should have a B.S. degree in science or engineering, two years experience in radiation protection, one year of which should be nuclear power plant experience, six months of which shold be onsite.
It is our position that such experience be pro-fessional experience.
Identify and provide an outline of the qualifications of the indivudual who will act as the backup for' the RPM in his absence.
I 471.06 In acccrdance with NUREG-0800, section 12.5, provide a resume of the (13.1 )
education, training, and experience of your Health Physics Superin-tendent.
471.07 In section 12.3.4.1 of the FSAR, you have stated that " Critically (12.3)
Monitors, as stated in 10 CFR 70.24 and Regulatory Guide 8.12, are not needed." Provide a commitment date, when an application for an exemption to 10 CFR 70.24(b) will be filed with the NRC. This exemption request should be filed as part of your application for V-a Special Nuclear Materials license.
'g 471.08 As requested in NUREG-0800, section 12.5 and 13.1, indicate how 2
(12.5) radiation protection training will be conducted for health physics professionals and health physics technicians at Vogtle. Radiation protection training programs for the levels of Health Physics tech-j nicians should be generally described to include initial qualifi-cation and retraining /requalification programs and should verify that selection, qualification and training requirements for con-tractor health physics technicians and contractor radiation workers are the same as or equivalent to the requirements for Vogtle radia-tion control technicians and radiation workers.
471.09 As required in Regulatory Guide 1.70, section 12.5.3, verify that a routine alpha monitoring program which includes routine conta-l mination, airborne and direct surveys for alpha will be conducted for Vogtle.
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471.1 0 Provide a discussion of the Radiation Protection Plan (RPP) intended (12.1) for Vogtle, as described in section 12.1 of your FSAR.
471.11
.Your portable radiation monitoring instrument list (table 12.5.2-2) and area radiation monitor list (12.3.4-1) show no instruments capable of measuring exposure rates greater than 1000 R/hr.
Such instruments are necessary to detemine the effects of a TMI-type accident. Regulatory Guide 1.97 (Revision 2) specified the area radiation monitors in areas requiring access after an accident and portable survey meters should have a range up to 10,000 R/hr.
Provide a commitment in your FSAR to have portable radiation moni-toring instruments and specify locations of area radiation monitors in accessible post-accident areas. These monitors should be capable of measuring Exposure rates up to 10,000 R/hr at Vogtle.
471.1 2 Section 12.5.3.9 of the FSAR states, " sealed radionuclide sources (12.5) having actjvities greater than the quantities of radionuclides defined in Appendix C of 10 CFR 20 and schedule B of 10 CFR 30 will be subject to material controls for radiological protection."
Since, (1) the radionuclides and activities listed in Appendix C are associated with allowable sewerage release limits authorized in 10 CFR 20.303, and not intended as dominimus quantities, and (2) sealed sources obtained under 10 CFR 30.18 may be redistributed only under a specific licensee issued in accordance with 10 CFR 32.18, you should revise your FSAR and procedures to require that all licensed sources be subject to material controls.
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, 471.1 3 Provide additional information regarding the sensitivity of air-l borne radioactivity monitors in accordance with section 12.3 of Regulatory Guide 1.70.
Verify that the airborne radioactivity
' monitors described in section 12.3 of the FSAR are capable of detecting 10 MPC-hours of particulate and iodine radioactivity in compartments which may be occupied and may contain airborne radioactivity (the acceptance criteria in Standard Review Plan section 12.3).
471.1 4 Section II.B.2 of NUREG-0737 requests that the applicant provide (12.3) the projected doses to individuals for necessary occupancy times in vital areas following an accident and dose rate zone maps for potentially occupied areas.
Figures 12.3.1-2 of the Vogtle FSAR show the radiation levels for various plant areas at twenty-four hours after an accient.
Provide a table giving the post-accident dose rates to vital areas for various times following an accident:
~.
e.g.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,1 day,1 week, and 1 month.
471.15
,Section II.F.1-3 of NUREG-0737 requests the applicant to provide (12.3) plant layout drawings showing the location of the containment monitors.
In addition, either provide the manufacture name and model number of these instruments or provide their specifi-cation.
471.16 The PWR exposure data used in Chapter 12.4 does not include plant (12.4) explosures after 1977. Update your expoure estimates to include more recent exposure information (NUREG-0713, Vol. 3 contains exposure data up through 1981).
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t 471.17 Section C.4.e.(4) of Regulatory Guide 8.8 recommends that change rooms be equipped with sufficient lockers to accomodate perman.ent and contract maintenance workers who may be required during major outages.
Show that your change and locker room facilities have this capability. Also, show that you have made provisions for both male and female employees.
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