ML19340A701
| ML19340A701 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/04/1961 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19340A698 | List: |
| References | |
| NUDOCS 8009030705 | |
| Download: ML19340A701 (15) | |
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<::rM COMMONWEAI/PH EDISON COMPANY GENERAL ELECTRIC COMPANY Proposed Amendment to License DPR-2 as Amended Changes to Appendix "A" By:
Commonwealth Edison Company General Electric Company
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- I'A January 1, 1961 4
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N INTRODUCTION Commonwealth Edison Company is operating the Dresden Nuclear Power Station under License DI'R-2, as amended, dated October 14, 1960, he re -
inafter referred to as "Ltcense DPR-2".
This current operation, as well as the prior test operations, have suggested the desirability of several changes or additions to the " technical specifications" for the reactor.
Appendix A to License DPR-2 sets forth the category of specifications that would be affected-by these proposed changes. This report, therefore, constitutes the technical information for an application to amend License DP R-2.
The changes proposed herein have principally to do with improving plant operability or convenience, corrections or clarifications to the current specifications, and improvement in the ability to demonstrate compliance with License DPR-2. The proposed changes embody few, if any, reactor safety considerations. Therefore, the " justification" which accompanies each proposed change is generally only a brief explanation of the background or reasons for the change.
This report is an addition to, but does not amend, other parts of Common-wealth Edison's license application. In any case where the information presented herein is not consistent with information in pre"iously subr.itted parts of the application, the information contained herein shall govern.
In cases where a change may involve changes in equipment, it is requested that any license amendment submitted pursuant to this application provide an overlap of 45 days from the effective date of the license amendment during which the provision of either the present license or the amended license may govern the operation of the Dresden Station. This will per-mit equipment changes to be made at a convenient time. Procedural and other changes may becon-' effective immediately, and no overlap is neces-sary for such changes tt lic ense.
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2-PROPOSED AMENDMENTS TO LICENSE DPR-2 1.
On page A-10, item "!4. Control Rods and Drives" of section "B. Design Features", delete in its entirety the last paragraph which reads:
"The drives are mounted on the bottom of the reactor vessel, and withdraw the control rods below the core. Upward movement of the control rods, into the core, decreases reactivity. The inserted position of the control rods is determined by a locking device which provides 12 discrete equidistant positions for all rods except those in the outer ring. The last withdrawal step for these is slightly shorter than the others".
and substitute therefor the following:
"The drives are mounted on the bottom of the reactor vessel, and withdraw the control rods below the core. Upward movement of the control rods, into the core, decreases reactivity. The inserted position of the control rods is determined by a locking device which provides 12 discrete approximately equidistant positions for all rods. "
II)
Justification:
The Dresden Technical Specifications report from which the present wording was taken, is incorrect in indicating that the outermost ring of control rods has a slightly shortened stroke. A change was made late in design to provide all control rods with the same stroke length. This information is presented II correctly in Amendment 3 on pages 10 and 50.
On all drives the length of withdrawal steps or " notches", is about
- 8. 8 inches except for the first notch which is about 8 inches.
2.
On page A-11, item "5. Liquid Poison System" of section d
"D. Design Featuren, delete in its entirety the entry which reads:
" Minimum poison worth 0.20 Ak"
_3 and substitute therefor the following:
" Minimum poison worth for operations with reactor vessel head on 0.15 Ak Minimum poison worth for operations with reactor vessel head off 0 04 Ak" Justification:
It is to be noted that there is no change in the mini-mum amount of boron contained in the poison system. Thus, there is no change in poison worth in the present system, and hence no change in the conservatism of the present system. However, the indicated poison worth of 0. 20 Ak was based on the minimum worth conditions for power operation, and does not necessarily represent the system, worth or need for worth under other conditions. It is recognized, for example, that use of the poison system under con-ditions of reactor refueling would initially provide a poison worth far in excess of 0. 20 Ak, but later when diluted with the substantial quantities of water in the refueling canal (with which the reactor water is in communication during refueling) would provide far less than 0. 20 Ak.
The proposed specification of 0. 04 Ak as the minimum poison worth for operations with the reactor vessel head off recognizes the worst possible dilution effects. The specified minimum worth of 0. 04 Ak is considered entirely adequate for situations which might occur when the reactor vessel head is oft because it is well in excess of the maximum worth of a single control rod (about O. 03 Ak) or the maximum worth of a single fuel assembly (about 0. 02 Ak). Also, dilution to the minimum value would not occur for some time, thus providing time for other provisions to be made if considered desirable.
The change in the specification for minimum poison worth with the reactor vessel head on, from 0. 20 Ak to 0.15 Ak, is being made at this time principally for administrative convenience. Since another
4 poison system specification is being added in the near future, and because the 0.20 A k specification would have to be changed in the near future if not changed now, this opportunity has been taken to make the change.
As stated earlier, the present posson system worth in the present core is still in excess of 0. 20 Ak. However, in some contemplated future core loadings, the worth may be reduced to somewhat below
- 0. 20 Ak due, for example, to the effects of.tdditional steel in the core. The new specification of 0.15 Ak represents a minimum worth which should be valid for all presently contemplated loadings. Also, 0.lSak poison worth is more than adequate for the present core since it exceeds the worth of the entire control rod system. Thus, the proposed reduction in the specification does not represent any change in the conservatism and high standards of safety maintained at the Dresden Station.
3.
On page A-12, item "6. Steam Supply System" of section "B. Design Features", add the following new paragraph to the end of item 6:
"The reactor recirculating pumps will be tripped whenever the level in the steam drum falls to approximately 21 inches below the drum center line. The pump trip is not a safety circuit function, but rather is provided for protection of the recirculating pumps against loss of inlet suction".
Justification:
In the present design, in order to protect the integrity of the recirculating pumps provision was made for tripping off such pumps several seconds after every scram in order to avoid pump cavitation.
This trip is accomplished by an auxiliary circuit off the safety system which includes a time delay relay. The purpose of this circuit was to provide normal flow for long enough after scram to permit a substantial part of the heat stored in the fuel to be transferred to the coolant, but still to trip the pumps before sufficient subcooling could be lost to cause pump cavitation. The experience with the system, however, has
-5 shown that cavitation is not a problem even if the pumps are not tripped off following scram. Therefore, the trip after every scram is to be replaced by a trip only upon the occurrence of a very low drum level, as indicated above. This trip will provide full protec-tion for the equipment and will improve plant operability by elimi-nating unnecessary delays now caused in having to restart the pumping loops after every scram. The pump trip would be preceded by a reactor scram at a higher drum level as indicated in Table I
" Safety System".
3 On page A-12, item "7. Main Condenser" of section "B. Design.
4.
Features", delete in its entirety the second paragraph, which reads:
"The following condenser vacuum trips will be operative except during 'Startup' or ' Shutdown' conditions:
Reactor scram if vacuum falls to or below 22 in Hg Turbine bypass valves close if-vacuum falls to or below 7 in Hg" and substitute therefor the following:
"In addition to the condenser vacuum scram and isolation trips indicated later in item 9 " Safety System", a mechanical trip is provided to close the turbine bypass valves, as well as all other hydraulically-opened turbine valves, should the condenser vacuum fall to or below 7 inches of Hg. This vacuum trip is not a part of the reactor safety system, but becomes operative when-ever the condenser vacuum has been increased to above the trip set point. "
Justification:
No changes in equipment or settings are involved in this proposed change. The present wording is somewhat in error and is in conflict with the information in item 9 " Safety System".
The change is made to correct and clarify the actual situation.
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-6 5.
On page A-16, Table I " Safety System", delete in its entirety the entry under " Scram trip setting" for the condition of " Low water level in primary steam drum" which reads:
"At level which is a maximum of 5" below the center line" and substitute therefor the following:
"At level which is a maximum of 12" below the drum center line".
Justification:
Based on the experience to date at the Dresden Station, 1
it would be desirable to increase the working range of the steam drum level to better accommodate such things as load changes and recirculation pump trips. At present, there is only about a 13-inch differential between the "high drum level turbine trip" *etting and i
the "lov, drum level reactor scram" setting. This relatively small working range can result in unnecessary turbine trips upon system
" swell" and unnecessary reactor scrams upon system " shrinkage".
The high level setting, based on steam separation requirernents, will be retained at its present setting. The low level scram setting was set originally to accommodate the calculated drop in level from instantaneous collapse of all system voids without emptying the steam dr.um.
However, test measurements of level shrinkage have indicated that only a 15-inch maximum drop in level occurs upon scram from full power. This 15-inch drop, iri conjunction with the proposed scram level setting at 12 inches below drum center line, will not empty the drum nor hinder recirculation flow. The recir-culation pump trip at 21 inches below drum center line would cause the pumps to be tripped in this case, but flow would be continued by natural circulation. Shrinkage following scrams from causes other than low drum level would not, in general, cause the pump trip point to be reached. Coincidentally with the change in low drum level scram setting, the rated power normal drum level will be reduced about three inches, to approximately divide the increased working range of 7 inches between margin against system swell and system shrinkage.
Since the low drum level scram, at its proposed new setting, will still fulfill its intendad function of preserving adequate drum level for proper reactor recirculation through the drum following scram, these proposed changes have.lo adverse effect on reactor safety.
. 6.
On page A-16, Table I " Safety System", delete in its entirety the entry under " Remarks" for the condition of " Low Condenser Vacuum" which reads:
" Bypassed in ' Shutdown' or ' Refuel' positions. Bypassed in
' Start' position if reactor pressure below 200 psig and condenser vacuum greater than 15" Hg. "
and substitute therefor the following:
" Bypassed in ' Shutdown' or ' Refuel' positions. Bypassed in
' Start' position if reactor pressure is below 200 psig and con-denser vacuum is greater than 10 inches of Hg. "
Justification: The sense of the change indicated above is to require that 10 inches of Hg vacuum be established with the mechanical vacuum pump before control rods can be withdrawn at startup, rather than 15 inches of Hg. There is no real safety significance in whether 10 inches or 15 inches of vacuum is used at startup. The safety significance of the interlock is only to assure that some vacuum has been established.
It has been found very difficult to maintain 15 inches of vacuum with the mechanical pump, but very easy to maintain 10 inches. Thus, it is for ease of operation that this change is requested.
Figure D-40 " Reactor Safety System Diagram" in Amendment 3( }
shows a setting of 10 inches of Hg for this interlock. As indicated by this drawing, it had been the intent to use a setting as low as the 10" Hg setting that is now requested. The 15" Hg setting, which was the upper limit of the planned 10-15 inch working range, was inadvert-II}
ently used in the Technical Specifications
, and has been adhered to in the operations to date, but the inconvenience it has caused makes it very desirable to return to the intended minimum value of 10 inches of Hg.
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. 7.
On page A-19, item "11. Reactor Enclosure" of section "B. Design Features", in the second paragraph on the page which reads:
"About one year after initial power operation, a leakage test of the enclosure will be conducted at a low pressure (les-than 10 psi) to determine the reliability of penetrations. (It is expected that this test will probably be made during the time the plant is down for the first major inspection of the turbine which is also expected to be done after about a year's operation. )"
delete the fifth word " initial" and substitute therefor the word "r e gula r".
Justification:
The original intent of the words " initial power opera-tion" was to convey the meaning more clearly expressed by the words
" regular power operation" which are now proposed to be used. This change is desired so there can be no confusion of the intended " initial" operation with operation when power was' initially" produced by the reactor in the test phases of operation prior to complete turnover of the plant from General Electric Company to Commonwealth Edison Company. As indicated in the parenthetical sentence, the intent was to schedule the leakage test at the time of the major inspection of the turbine, which would be after about one year's operation of the turbine by Commonwealth Edison. The proposed change in wording also con-forms to the language of testimony on this same subject to be found on page 996 of the Transcript of the September 26, 1960, Public Hearing concerning the Dresden full-term operating license.
Inasmuch as there is no safety significance in whether the test is conducted several months sooner or later than indicated, there is no change in plant safety whicheve. Enterpretation of the words is used. The change is requested to clarify the commitment actually made by Commonwealth and to provide consistency with the rest of the public record on this matter.
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8.
On page A-19, item c. of section "C. Operating Principles" add the,following:
"except when the safety system selector switch is in the shut down position and locked.
During this time one licensed operator will man the control room."
Justification:
When the four-position safety system selector switch is in the " shut down" position the control rods cannot be withdrawn.
As this switch will be locked in this position, it cannot be accidentally moved to another position thereby making it possible to move a centrol red.
Under this condition there is no need for a second man in the control room.
9.
On page JL-22, delete in its entirety item 4 " Shutdown Margin" which reads:
"4.
" Stuck. Rod" Criterion:
At every stage during the initial loading, and in the fully loaded configuration, the control rods must provide a shutdown control margin of at least O. 01 Ak with any rod wholly out of the core and completely unavailable.
b.
" Cocked Rod" Criterion:
During core alterations after the first fuel cell is loaded, the reactor must be subcritical by at least 0. 01 Ak with at least one control rod fully withdrawn in the region of the alteration and available for rapid scram insertion. "
and substitute therefor the following:
"4.
" Stuck Rod" Criterion:
The control rods must provide a shutdown control margin of at least 0. 003 Ak with any control rod fully withdrawn and any adjacent control rod withdrawn to the first notch. This margin must be pro-vided in the fully loaded core configuration and in any smaller configuration capable of sustaining criticality.
The 0. 003 Ak margin is to be demonstrated using the falling period technique.
In applying this criterion in cases where some centrol rod or rods may be known to be " stuck" or inoperable for any reason, one or more of the following additional criteria shallapply:
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Any stuck or inoperable control rod which has been taken out of service may be regarded as fixed permanent poison, and the " stuck rod" criterion must be demonstrated using the remaining operable control rods. Thus, a fully inserted " stuck" rod will not cause any difficulty in meeting the stuck rod criterion, but a partially or fully withdrawn stuck rod may cause difficulty in some parts of the core. If the stuck rod criterion cannot be met j
by some control rods because of a rod which is stuck out, it will be permissible to insert an other-wise-operable rod (or rods) to such a position that the criterion can be met by the remaining operable rods. An otherwise-operable rod which is inserted in this fashion as fixed permanent poison must be taken out of service and be regarded the same as a i
stuck rod.
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2)
A control rod is considered as "taken out of service" a
1 when it is mechanically locked in a discrece position by the collet fingers or is fully withdrawn, the drive's movement actuation valves have been dis-connected electrically, and the manual valves in series with the movement actuation valves have been closed manually and tagged.
Cooling water will still be provided for the drive.
3)
In no case shall the reactor be brought critical or 2
remain critical with a control rasi stuck in the core and separated from its drive mechanism in such a way that the rod would fall if it became unstuck.
If the rod and mechanism are in contact, the unit may be treated as a " stuck or inoperable" control rod so far as meeting the stuck rod criterion is concerned.
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A control rod and drive which is stuck between notch positions shall be regarded the same as a stuck control rod which is separated from its drive mechanism.
5)
The conditions for continued power operation with rods which become stuck during operation is given in item E. 5. c.
b.
" Cocked Rod" Criterion:
During core alterations, whenever the core would be critical with all control rods withdrawn, at least one control rod in the vicinity of the alteration will be withdrawn approximately half-way and will be maintained available for rapid scram I
inse rtion. "
Justification:
The present " stuck rod" criterion has been, in principle, a very meaningful safety criterion for the reactor.
This criterion is called the " stuck rod" criterion because it assures that the reactor can be shut down completely in the cold clean condition with any rod stuck wholly out of the core.
If such a case should occur, the core would still maintain a minimum shutdown margin of 0.003 A k plus the worth of the strongest first notch of any control rod adjacent to the stuck one.
However, this criterion has proved difficult to apply in practice.
A rather time consuming procedure has been required in order to demonstrate compliance with the " margin of at least 0.Ola k" part of the criterion.
Also, inaccuracies in the technique of measurement used have tended to make the actual minimum margin more than 0.014 k in order to assure compliance.
The proposed " stuck rod" criterion is one which is believed to be quite comparable with the present one from the safety standpoint, and should be very simple to apply from the operational viewpoint.
In practice, compliance with this criterion would be demonstrated by fully withdrawing the most limiting control rod, withdrawing the most limiting adjacent rod one notch (about 8 inches) and observing a falling period of not more than 80 seconds (corresponding to a e
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. i negative Ak of at least O. 003) when all other control rods are inse rted. Sufficient "most limiting" rods willbe checkedin this fashion to assure that the stuck rod criterion can be met in t
all regions of the core. The many tests made during initial loading of the Dresden reactor provide considerable guidance as to which rods are likely to be most limiting.
In the first Dresden core the most limiting shutdown margin conditions were to be found at the core periphery. For this region of the core, it is estimated that the strongest first notch adjacent to a withdrawn control rod is worth at least 0. 002 Ak. Combined with the 0. 003 Ak margin which is measured by the falling period technique, a shutdown margin of at least 0. 005 Ak will be maintained even if the strongest control rod should stick in the fully withdrawn position. The notch worth may vary considerably with exposure conditions and with future reloading patterns, but it is not anticipated that the notch worth of interest in the proposed " stuck rod" criterion would ever be so large as to impose a more restrictive criterion than the present "0. 01 Ak" c rite rion. Thus, the proposed criterion may be slightly less conserva-tive than the present criterion. However, this difference in conserva-tism is small, the proposed criterio.1 is believed to be very adequately safe, and the proposed criterion more than makes up for any small reduction 'u conservatism by its much greater ease of application.
The added criteria on how rods which are actually stuck are to be treated in determining compliance with the " stuck rod" criterion are merely interpretations of the basic criterion. They are set forth to clarify the intent of the basic criterion for some specific cases and to put into the technical specifications some information submitted in a previous report.
The proposed change in the " cocked rod" criterion eliminates the "0. 01 Ak" part of the present criterion, which is a desirable change for the same reasons as discussed for the " stuck rod" criterion. It a
4 is to be noted that the proposed " cocked rod" criterion does not specify any numerical limits on " cocked rod" worth or on the minimum shutdown margin with rods cocked. It is desirable that these details be left to the discretion of the operating personnel, inasmuch as the different circumstances encountered in various core locations permit optimum safety to be achieved with different amounts of reactivity held in cocked safety. For example, a core region which is normally shut down by 0. 05 Ak would require a different numerical " cocked rod" criterion than a region normally shut down by only O. 03 Ak.
The proposed " cocked rod" criterion requires the use of at least one cocked rod. Actually, except when loadir in the periphery, two cocked rods are used during core alterations. In the periphery it is not possible to use two cocked rods without encountering at least one of the following difficulties:
a.
Holding too much of the shutdown margin in cocked safety.
i b.
Having at least one of the cocked rods sufficiently removed from the location of the core alteration that it has insignifi-
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cant reactivity worth.
c.
Unduly enhancing the worth of an inserted control rod or the worth of a fuel bune ie being inser ed into the region being altered.
The use of one cocked rod for peripheral loading and two cocked rods for other regions of the core follows the practice actually used at the Dresden Station to date.
The. fact that the cocked rods must be in the vicinity of the alteration assures that the ~ rods will hdv'e "significant" worth in that region.
Flexibility in deciding which rods to use in the " vicinity" of the e
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.. alteration permits having "significant" cocked rod worth without cocking so much of the shutdown margin as to increase the likeli-hood of reaching accidental criticality during the core alteration.
In general, not more than about half of the estimated minimum shutdown margin for the region would be held in cocked safety rods. The use of approximately half-withdrawn rods is intended to improve the scram response to accidental criticality even if it occurs near the top of the core region of interest.
10.
On page A-24, item "2. Safety System Scram Settings" of section "E. Power Operation", delete in their. entirety the words "15 inches of Hg condenser vacuum" on the last line of the page, and substitute therefor the words "10 inches of Hg condenser vacuum".
Justification:
See change number 6, above.
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REFERENCES (1)
GEAP-3082, " Summary and Technical Specifications for the Dresden Nuclear Power Station", by J. L. Murray and D. P. Ebright, February 3,1959.
(2)
GE AP-3076, " Amendment No. 3 to Preliminary Hazards Summary Report for the Dresden Nuclear Power Station",
by J. L. Murray and D. P. Ebright, Decembe r 23, 1958.
(3)
" Safeguards Against a ' Stuck Rod' Incident", by General Electric Company, March 8,1960.
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