ML19257C430
| ML19257C430 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/16/1980 |
| From: | Holt D SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | CALIFORNIA, STATE OF |
| Shared Package | |
| ML19257C422 | List: |
| References | |
| NUDOCS 8001290110 | |
| Download: ML19257C430 (25) | |
Text
.
A January 16, 1980 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY DISTRICT
)
Docket No. 50-312
)
(Rancho Seco Nuclear Generating Station)
)
LICENSEE'S ANSWERS (SET NO. 3) TO CALIFORNIA ENERGY COMMISSION'S SECOND SET OF INTERR0GATORIES TO THE SACRAMENTO MUNICIPAL UTILITY DISTRICT 1.
INTERROGATORY:
Identify each transient contemplated in subparts (a),
{b), and (c) of SMUD's Answer to CEC Interrogatory No. 5 (i.e., which transients may result in reactor coolant pressure reduction below the HPI setpoint or may block use of AFW) and CEC Interrogatory No. 6 (i.e.,
which transients may result in turning off the main coolant pumps) of CEC's First Set of Interrogatories to SMUD, including the following:
(a)
The likely initiating event; (b)
The expected frequency of the event; (c)
The cause of the resulting RCS pressure loss (if applicable);
(d)
The reason the reactor coolant pumps would not operate (if applicable);
(e)
A schedule of all events that SMUD expects will folicw initiation of the transient, including all operator or plant system responses and their expected effects, and the timing of each event; (f)
A schedule, as per subpart (e), above, of the events SMUD expects will follow the initiation of the transient if it is initiated or accompanied by a loss of of f-site power; (g)
If PORV actuation is expected (either with or without a loss of off-site power), estimate how long the PORV will remain open and the total volume of coolant that would be lost from the primary i85fr265 system; and (h)
Provide primary system temperature and pressure profiles for each identified type of transient.
ANSWER:
The list of plant transients which require or may require the use of high pressure injection to assure core cooling, and those which require or may require the use of natural circulation for core cooling is provided in the attached Tables lA and 1B, respectively.
Tha information requested in subparagraphs (a) through (h) for each of the transients is also given k
8001290
in Tables lA and IB. The list of transients identified is based on events which have occurred at Rancho Seco; other (postulated) transients as identified in Chapter 14, " Safety Analysis", of the Rancho Seco FSAR; and evaluations perfonr.ed since the TMI-2 incident. For a discussion of the terms used with regard to frequency in Tables l A and IB refer to Standard ANSI N18.2-1973,
" Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants". A copy of this document will be provided in response to the
" California Energy Commission's Second Request for Production of Documents to the Sacramento Municipal Utility District".
2.
INTERROGATORY:
Explain any sensitivities of B&W reactor systems that may require use of HPI in conjunction with feedwater transients.
[.SWER:
Durir.g anticipated feedwater transients the design of the B&W nuclear steam system at Rancho Seco does not require use of high pressure injection.
If a severe overcooling transient occurs (e.g., loss of main feedwater followed by excessive addition of auxiliary feedwater) high pressure injection may be required to assure adequate core cooling. Such events, one of which has occurred at Rancho Seco, are expected to be of low frequency.
They involve failure or maloperation of the steam generator cooling or steam relief systems, and are not unique to the B&W design.
3.
INTERROGATORY:
In the event of a feedwater transient, will reactor pressure always be maintained above the HPI setpoint?
If not, how often do you expect the HPI to be activated?
Is this more frequent than with non-8&W systems?
ANSWER:
See answer to Interrogatory 2.
1 b tJr 2 6 6 6.
INTERROGATORY:
Identify the safety systems challenged by an increase in reactor trips and compare the design basis of each such system with anticipatea challenges.
(
Reference:
Description of testimony of Mr. B. A. Karrasch wi+'
regard to CEC Issue 1-12),
Af.5WER:
The safety systems expected to be affected (reactor coolant, reactor protection and control rod drive) are designed for 400 reacar trip cycles.
Prior to the TMI-2 accident, the reactor trip frequency for.
B&W units was below the industry average. With the additional reactor trips now anticipated, the trip frequency for the B&W units is expected to increase and approximate the industry average. Applying this average frequency for Rancho Seco, the limit of 400 reactor trip cycles would not be exceeded. The actual number of design transients is monitored -
see " Licensee's Answers (Set No. 2) to the California Energy Commission's First Set of Interrogatories Dated November 15, 1979", Interrogatory 30, Issues CEC-1-1 and CEC-1 to assure that the design limit is not exceeded.
11.
INTERROGATORY:
In the limiting circumstance where auxiliary feedwater would normally be required but such is not available, describe how adequate core cooling is maintained. What procedures and operator actions are necessary in dealing with such events? Can the NSSS be placed into stable, long-term cooling without auxiliary feedwater?
ANSWER:
In the limiting circumstance where auxiliary feedwater would normally be required but such is not available, adequate core cooling can be maintained by actuation of high pressure injection and energy release via the pressurizer relief and/or safety valves. This is accomplished by manual initiation of high pressure injection and, if utilized, opening of the pressurizer relief valve (PORV).
The nuclear steam system can be placed in a stable long-term cooling mode without auxiliary feedwater.
12.
INTERROGATORY:
Under I&E Bulletin 79-05C, the primary coolant pumps must be tripped when the HPI is activated.
(a)
Describe the location and size of the small breaks which prompted the 79-05C procedure. Why are the pumps turned off? What may happen if the pumps are left on?
(b)
Are there circumstances where the HPI is on but restart of the primary pumps may be required?
(c)
How is adequate core cooling maintained?
(d)
Describe the instruments and controls available to 1o 9gf an operator to insure fuel rod integrity in the event iU CU/
that water inventory and/or cooling is marginal.
What procedures have been developed ant are available to operators under these circumstances?
ANSWER:
(a)
With regard to NRC IE Bulletin 79-05C, t a relevant small breaks have been identified as those, regardless of location, ranging in size 2
2 fram 0.025 ft to 0.2 ft. For these breaks, the reactor coolant pumps are shut off because continuous circulation of reactor coolant results in a high system void fraction and, if the pumps are shut off af ter a high reactor coolant system void fraction is reached, the decreased system inventory may result in inadequate core cooling.
If the reactor coolant pumps remain in operation throughout the transient, the continued forced circulation will provide adequate core cooling.
(b) Restart of the reactor coolant pumps is not required to assure adequate core cooling. However, operating guidelines may direct restart of the pumps under certain circumstances - for example, long-term management of plant conditions.
(c) For the subject small breaks, following trip of the reartor coolant pumps as specified, adequate core cooling is maintained by the emer-gency core cooling systems.
(d) Reactor coolant system temperature and incore thermocouples are available to the operator to assess core conditions in the event cooling is postulated to become marginal.
Operating guidelines have been developed for inadequate core cooling which provide for restart of a reactor coolant pump (or pumps), depressurization of the steam generators, opening of the pressurizer relief valve and/or initiation of reactor coolant system makeup.
13.
INTERROGATORY:
If a small break exceeds the capacity of HPI, are there conditions whereby the core could be uncovered before othe. ECCS components would be expected to operate? Provide copies of any analyses which relate to your response.
ANSWER:
Small breaks in the reactor coolant system in excess of approximately 2
0.04 ft require operation of ECCS components other than high pressure injection, when analyzed in compliance with 10CFR50, Section 50.46.
The analyses referenced in a July 18, 1978 letter from J. J. Mattimoe (SMUD) to R. W. Reid (NRC) demonstrate that adequate core cooling is assured although partial core ifda.268 uncovery is indicated. A copy of this anaylsis will be provided in
_a_
response to the " California Energy Commission's Second Request for the Production of Documents to the Sacramento Municipal Utility District."
14 INTERROGATORY:
Provide the basis of the following statement, which appears at p. 6 of SMUD's Answers (Set No. 3) to the First Set of Interrogatories of the CEC.
"For anticipated loss of feedwater transients, void formation does not occur in the B&W nuclear steam system."
ANSWER:
For Rancho Seco as currently operating, an anticipated loss of main feedwater transient can tie characterized as follows:
Loss of main feedwater occurs.
Reactor coolant temperature decreases due to the loss of heat source (reactor power) and stabilizes at the saturation temperature for the steam generator secondary side pressure.
Pressurizer level initially decreases due to reactor coolant volume contraction, then returns to the normal control setpoint in response to increased makeup flow.
Reactor coolant pressure initially decreases due to reactor coolant volume contraction, then returns to the normal control setpoint in response to the action of the pressurizer heaters and increased makeup flow.
Steam generator secondary side level initially decreases due to the loss of feedwater addition, then is maintained at a preset level by auxiliary feedwater flow.
Secondary side steam pressure is controlled by the turbine bypass system.
Throughout the transient the pressurizer does not empty and the reactor ]gg }gg coolant remains subcooled, therefore there is no void formation. -
15.
INTERROGATORY:
Were conditions of core voiding and loss of natural and forced circulation cooling considered during the licensing of Rancho Seco? If so, discuss.
ANSWER:
Yes.
For example, large break less-of-coolant-accident analyses for Rancho Seco indicated conditions of core voiding and assumed no natural or forced circulation cooling.
Refer to Section 14.2.2.5 of the Rancho Seco Final Safety Analysis Report.
16.
INTERROGATORY:
How is water inventory maintained under the conditions of the largest size break not requiring low pressure injection and for which significant core voiding has occurred such that density head natural circ-ulation does not perform? Can adequate core cooling be maintained where such conditions are accompanied by a loss of off-site power and one diesel genera tor? Explain.
ANSWER:
The largest break in the reactor coolant system not requiring low pressure injection has been identified as a 0.44 ft break in a core flood tank line. Water inventory for this break is maintained by high pressure injection and the other core flood tank. Assuming loss of off-site power and one diesel generator, mitigating equipment available would be one high pressure injection train and one core flood tank.
These systems provide sufficient injection to assure adequate core cooling.
Natural circulation is not required.
17.
INTER,ROGATORY:
Will the steam cenerators be available for core cooling during or following a severe feedv3 ?er transient? Explain. What measures have been taken to limit the potentul for steam generator failure due to steam bubble collapse, tube failure cteated by thermal shock, or other known cause for potential failure?
ANSWER:
Yes. Analyses have been performed assuming a conservative set of primary and secondary conditions following a severe feedwater transient such as to maximize steam generator tube thermal stress. The results showed that the integrity of the steam generator tubes would be maintained, and thus a breach of the primary pressure boundary would not be anticipated.
13.
INTERROGATORY:
Describe and discuss the Rancho Seco reactor coolant systen configuration and the adequacy of the driving head for natural circulation.
Describe those analyses which provide a basis for any conclusions regarding the adequacy of maintaining core cooling including methodology, conclusions, assumptions and uncertainties assocated with these analyses.
AllSWER:
The Rancho Seco (B&W) reactor coolant system has been proven to be capable of providing adequate circulation by the following:
Analyses have been performed to determine that natural circulation is adequate to maintain core cooling when required due to all reactor coolant pumps being inoperative.
These analyses were performed over a wide range of plant conditions, utilizing conservative assumptions.
flatural circulation testing has been conducted at B&W operating plants. The testing confirmed that natural circulation can be initiated and maintained over a wide range of plant conditions, and demonstrated that the design analyses conservatively predict the natural circulation capabilities of the plants.
Operational occurrences of natural circulation core cooling have been experienced at B&W operating plants to further demonstrate the adequacy of the B&W system under this condition.
In all of these unplanned events, in which all reactor coolant pumps were inoperative, natural circulation maintained the plant in a safe condition.
Further explanation of the above is provided in Appendix 1 (Revision 1),
"flatural Circulation in Operating B&W Plants," to a May 7, 1979, R&W report entitled " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant"; and in a " Report on Analysis Methods for RCS tiatural Circulation" attached to a June 7,1979 letter from R. E. Ham (B&W) to the B&W owners.
Copies of these documents will be provided in response to the " California Energy Commission's Second Request US 27 for Production of Documents to the Sacramento Municipal Utility District."
TABLE 1A POSTULATED TRAllSIENTS REQUIRIrlG HIGH PRESSURE IrlJECTIO!1 1.
Excessive Feedwater Addition Transient (a)
Initiating Event:
Feedwater addition rate not properly controlled to match primary heat removal demand following a reactor trip.
(b) Frequency:
Moderate frequency (c) Cause of RCS Pressure Loss:
Contraction of primary coolant because of excessive heat removal due to overfeeding steam generator.
(d) Reason RC Pumps Would Not Operate:
RC pumps would operate except in the case of a loss of offsite power or a manual trip.
(e) Sequence of Events (Offsite Power Available):
- Reactor trip and turbine trip.
- RC pressure and temperature start to decrease due to reduction in core power output.
- Malfunction or maloperation of feedwater control system results in continued excessive feedwater flow to steam generators.
- Pressurizer empties due to primary coolant contraction.
- Low RC pressure engineered safety features actuation signal (ESFAS) initiates closure of main feedwater isolation valves and actuates high pressure injection system. Auxi lia ry feedwater actuated.
- Pressurizer starts to fill due to high pressure injection action.
Cperator controls HPI flow to reestablish and maintain pressurizer level and RC pressure.
1824 272 33.,
TABLE iA (cont'd.)
- Core remains adequately subcooled throughout transient.
(f) Sequence of Events (With Loss of Offsite Power):
The sequence of events for this transient is similar to (e) above.
The excessive feeds ater addition transient with RC pumps operating is a more severe ovi rcooling transient since operation of the RC pumps will incre.sse the primary heat transfer rate (cooldown rate) of the primary system.
(g) PORV Actuation:
PORV actuation is not expected.
(h) Primary System Pressure and Temperature Profiles:
See Figures 1.1 and 1.2.
18M 273 1A-2
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TABLE 1A (Cont'd) 2.
Steam Line Failure (a)
Initiating Event:
Double-ended steam line rupture.
Note:
A double-ended steam line rupture with concurrent failure of steam stop valve in unaffected steam generator has also been analyzed (see Rancho Seco FSAR). For that case, the primary system response is very similar and, therefore, not addressed here as a separate accident.
(b) Frequency:
Limiting fault (c) Cause of RCS Pressure Loss:
Contraction of primary coolant due to rapid overcooling caused by secondary side blowdown.
(d) Reason RC Pumps Would Not Operate:
RC pumps would opera +.e except in the case of a loss of offsite power or a manual trip.
(e) Sequence of Events (Offsite Power Available):
- Double-ended steam line rupture.
- Reactor trip on high flux or low RC pressure. Turbine trip and turbine stop valve and feedwater control valves start to close.
- Rapid decrease in secondary side steam pressure increases heat transfer from the reactor coolant to the steam generator feed-water which results in a rapid reduction in reactor coolant pressure and temperature.
- Low steam line pressure initiates automatic feedwater isola ~ tion which allows the steam generator associated with the rupture to blow dry.
183& 276
- Pressurizer empties.
- Low RC pressure ESFAS also initiates closure of main feedwater lA-5
TABLE 1A (Cont'd.)
isolation valves and actuates high pressure injection system.
Auxiliary feedwater actuated to supply feedwater to isolated, unaffected steam generator.
-Pressurizer starts to fill due to high pressure injection action. Operator controls HPI flow to reestablish and maintain pressurizer level and RC pressure.
-Core cooling is maintained with the aid of auxiliary feedwater flow through uneffected steam generator.
(f) Sequence of Events (With Loss of Offsite Puwer):
The sequence of events for this accident is similar to (e) above.
The st:3m line rupture accident with RC pumps operating is a more severe osercooling accident since operation of the RC pumps will increase the primary heat transfer rate (cooldown rate) of the primary system.
(g) PORV Actuation: PCRV actuation is not expected.
(h) Primary System Pressure and Temperature Profiles:
See Figure 2.1.
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1830 277
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TABLE 1A (Cont'd.)
3.
Steam Generator Tube Failure (a)
Initiating Event:
Complete severance of one steam generator tube.
(b) Frequency:
Limiting fault (c) Cause of RCS Pressure Loss:
Reduction in primary system inventory through failed steam generator tube.
(d) Reason RC Pumps Would Not Operate:
RC pumps would operate except in the case of a loss of offsite power or a manual trip.
(e) Sequence of Events (Offsite Power Available):
- Complete severance of a steam generator tube.
- Initial leak rate exceeds normal makeup to the reactor coolant system, and system pressure decreases.
- Reactor trips on low RC pressure. Turbine trips.
- Reactor coolant pressure continues to decrease until high pressure injection is automatically actuated. The capacity of the HPI system is suf ficient to compensate for the leakage and maintains volume control of the reactor coolant system.
- Within 20 minutes following the accident, the operator initiates a reactor coolant system cooldown and depressurization.
- When the reactor coolant system has been depressurized below the low steam safety valve setpoints, the operator completes isolation of affected steam generator and continues cooldown and depressurization of the reactor coolant system using the unaffected steam genera. tor for heat removal.
18jW 279 (f) Sequence of Events (With Loss of Offsite Pcwer):
lA-8
TABLE 1A (Cont'd.)
- The sequence of events for a steam generator tube failure accident coincident with a loss of offsite power would be similar except that core cooling would be achieved through natural circulation rather than forced circulation.
(g) PORV Actuation:
PORV actuation is not expected.
(h) Primary System Pressure and Temparature Profiles:
Not available.
18jVk280 1A-9
TABLE 1A (Cont'd.)
4.
Letdown Line Rupture (a)
Initiating Event:
Complete severance of letdown line.
(b) Frequency:
Infrequent incident (c) Cause of RCS Pressure Loss:
Reduction in primai, system inventory thrcugh ruptured letdown line.
(d) Reason RC Pumps Would Not Operate:
RC pumps would operate except in the case of a loss of offsite power or a manual trip.
(e) Sequence of Events (Offsite Power Available):
- Complete severance of letdown line.
- Leak rate tuceeds normal makeup to the reactor coolant system, and system pressure decreases.
Reacter trips on iow pressure. Turbine trips.
- Reactor coolant pressure continues to decrease until high pressure injection is automatically acti:ated on low RC pressure.
ESFAS action also initiates closure of letdown isolation valve.
- RC system inventory loss is terminated with closure of letdown isolation valve.
RC pressure is restored via high pressure injection and maintained through operator control of HPI flowrate.
(f)
Sequence of Events (With Loss of Offsite Power):
The sequence of events for a letdown line rupture coincident with a loss of offsite power would be similar except that core cooling would be achieved through natural circulation rather than f r ed 18 N 281 c i rc u la ticn.
(g) PORV Actuation:
l A-10
TABLE 1A (Cont'd.)
PORV actuation is not expected.
(h) Primary System Pressure and Temperature Profiles:
flot available.
1816 282 lA-11
TABLE 1A (Cont'd.)
5.
Small Break Loss of Coolant Accident (a)
Initiating Event:
Small break in the primary system greater than makeup system capability.
(b) Frequency:
Infrequent incident (c) Cause of RCS Pressure Loss:
Reduction in primary system inventory through break.
(d) Reason RC Pumps Would Not Operate:
RC pumps would operate except in the case of a loss of offsite power or manual trip.
(e) Sequence of Events (Offsite Power Available) - Typical:
Small break in primary system.
Leak rate exceeds normal makeup to the reactor coolant system, and system pressure decreases.
Reactor trips on low RC pressure. Turbine trips.
Low pressure ESFAS actuation initiates HPI and duxiliary feedwater, and closure of main feedwater isolation valves.
Operator manually trips RC pumps.
Operator controls level in steam gen'erator to 957, on operating range using auxiliary feedwater.
RCS reaches saturation.
Core flood system actuates (depends on break size).
Low pressure injection system actuates (depends on break size).
(f) Sequence of Events (With Loss of Offsite Power):
The sequence of events for a loss-of-coolant accident coincident with a loss of offsite power will be similar, except that the loss of offsite power will result in an immediate trip cf all 1836 283 1A-12
TABLE 1A (Cont'd.)
RC pumps which will also automatically actuate auxiliary feedwater.
(g) PORV Actuation:
PORV actuation is not expected during a loss-of-coolant accident.
For a limited range of breaks the PORV may be used by the operator to depressurize the RCS if natural circulation is interrupted and RC pumps are not available.
(h) Primary System Pressure and Temperature Profiles:
Typical primary system pressure responses are shown in Figure 5.1.
It should be noted that the primary system pressure response for a small break is highly dependent on the break size and location. System temperature will generally be at the saturation temperature corresponding to the system pressure during the transient.
183% 284 1 A-13
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lA-14
TABLE IB POSTULATED TRANSIENTS REQUIRING NATURAL CIRCULATION 1.
Loss of Coolant Flow (a)
Initiating Event:
Loss of power to all four RC pumps.
(b) Frequency Moderate frequency (c) Cause of RCS Pressure Los:,:
RC system pressure will decrease initially due to the reduction in core power immediately following the reactor trip but will stabilize as the secondary heat removal rate is reduced to match
.ne core decay heat level. Low pressure ESFAS will not be actuated.
(d) Reason RC Pumps Would Not Operate:
Loss of RC pumps is the initiating event.
(e) Sequence of Events (Offsite Power Available):
- Los'. of power to all four RC pumps.
- Reactor trip initiated on loss of all four RC pumps. Auxiliary feedwater initiated. Turbine trips.
- Steam generator levels increased by auxiliary feedwater to provide for natural circulation cooling for core decay heat removal.
(f) Sequence of Events (With Loss of Offsite Power):
This is addressed in Transient 2, Table IB.
1 b$'b 2b b (g) PORV Actuation:
PORV actuation is not expecteu.
(h) Primary Systim Pressure and Temperature Profiles:
Not available.
18-1
TABLE 1B (Cont'd.)
2.
Loss of All Unit A-C Power (a)
Initiating Event:
Loss of all A-C power.
(b) Frequency:
Moderate frequency (c) Cause of RCS Pressure Loss:
RC system pressure will decrease initially due to the reduction in core power immediately following the reactor trip but will stabilize as secondary heat removal rate is reduced to match the core decay heat level.
Low pressure ESFAS will not be actuated.
(d) Reason RC Pumps Would Not Operate:
Loss of all A-C power will cause removal of power to RC pumps.
(e) Sequence of Events (Offsite Power Available)
Not applicable (f) Sequence of Events (With Loss of Offsite Power):
- Loss of all A-C power results in RC pump loss, reactor trip, turbine trip, closure of turbine stop valves and initiation of auxiliary feedwater.
- RCS temperature and pressure are controlled by controlling secondary steam relief via main steam safety valves and atmospheric dump valves (if available).
- Steam generator levels increased by auxiliary feedwater to provide for natural circulation cooling for core decay heat remo (g) PORY Actuation:
PORV actuation is not expected.
(h) Primary System Pressure and Temperature Profiles:
See Figure 2.1 1 B-2
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMf11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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SACRAMENTO MUNICIPAL UTILITY DISTRICT
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Docket No. 50-312
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(Rancho Seco Nuclear Generating Station)
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AFFIDAVIT OF D. C. HOLT County of Sacramento)
SS State of California )
D. C. Holt, being duly sworn according to law, deposes and says that he is a Senior Engineer in the Nuclear Power Generation Division of the Babcock and Wilcox Company; and that the answers contained in
" Licensee's Answers (Set No. 3) to California Energy Commission's Second Set of Interrogatories to the Sacramento Municipal Utility District" are true and correct to the best of his knowledge and belief.
D. C. Holt Sworn to and subscribed before me this day of January, 1980.
Notary Public My Commission expires
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