ML19343B131
| ML19343B131 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 08/31/1962 |
| From: | Bryan R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19343B129 | List: |
| References | |
| NUDOCS 8012090239 | |
| Download: ML19343B131 (16) | |
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DATE: August 31, 1962 HAZARDS ANALYSIS BY THE RESEARCH AND POWER REACTOR SAFETY BRANCH
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DIVISION OF LICENSING AND REGULATION g:.
IN THE MATTER OF
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YANKEE ATOMIC ELECTRIC CO.
DOCKET NO. 50-29
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INTRODUCTION By letter dated June 4,1962, Yankee Atomic Electric Company submitted Amendment No. 41 to its license application and requested authorization to increase the~ maximum steady. state power level of the Yankee reactor from 485 to 540 Mw thermal. As a part of this request, Yankee furnished proposed superseding and additional pages which would be incorporated into the Yankee
" Final Hazards Summary Report" if' authorization to increase the power level were granted. These pages describe the changes which would be made in the facility design and operation in conjunction with the proposed reactor. power lm.:.....::Nb increase and analyze the effects that the increased power level would have
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on safety. Xt should be noted that part of the analysis presented by Yankee p
is based on operation at:580 instead of 540 Mw thermal operation..:The former F
figure is the power level for which Yankee had intended to request authoriza-
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tion; however, the results of their analysis of the effects on DNB (departure from nucleate boiling) ratios of increasing the maximum power level led to their requesting the lower value of 540 Mw thermal.
II.
BACKGROUND
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Following an initial period of operation of the Yankee reactor at 392 Mw thermal, Yankee Atomic Electric Company requested and was authorized an in-crease in the maximum steady state power level of their reactor to 485 Mw
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thermal. Operation at the higher power level commenced in June,1961 and continued, except for interruptions most of which were incidental to inspec-tions, modifications and tests, until December,1961. Following that date,
L kNb the maximum operating power was gradually decreased to increase reactivity and, thereby, extend core life. By May of this year, the power level of i
the reactor had declined to approximately 290 Mw thermal; and it was no longer economically practicable to continue operation. Accordingly, the reactor was shut down for refueling. The Core I fuel, except, for two as-semblies which are to remain in the reactor during Core II life for test purposes, was then replaced. In addition, the control rod drive shafts and coupling assemblies, as well as the control rod absorber sections and fol-lovers, were replaced with ones of a new design.
During Core I life and the subsequent core refueling operation, Yankee obtained considerable data and experience. The reactor in-core instrumenta-
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tion provided data from which a number of Core II parameters can be estimated with considerable assurance, especially since Yankee plans to use the same control rod program and essentially the same fuel loading during operation with Core II.
Experience gained during Core I operation has also resulted
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in a number of minor changes in the f acility design and mode of operation.
It is our opinion that these changes should contribute to the safety of operation of the facility.
As would be expected in a large power reactor facility such as Yankee, a number of problems have been encountered, such as:
A.
Anomalous reactivity changes were encountered during operation following shut down after power operation. Such changes were s
also encountered after the pH of the main coolant water was in-creased to determine what effect it might have on core reactivity.
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Wear occurred on certain parts of the control rod assemblies at the juncture between the control rod drive' shafts and the absorber sections and between the absorber sections and the followers.
C.
Deterioration of the dif fusion bonded Ni cladding on the Ag-In-Cd control rod absorber sections has been experienced.
D.
High levels of radiation were present in the area above the shield tank cavity when it was filled with water during the refueling operation at the end of Core I life.
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We believe, however, that these problems have not resulted in. the reactor being operated in an unsafe mannerrand that it has been operated without undue hazard to the health and safety of the public. It is also our
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opinion that these problems have little or no significance with respect to the increase in power.
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III. CRANCES IN FACILITY TO BE MADE PRIOR TO INCREASING POWER LEVEL Yankee proposes to make a number of changes in the design and operation of the facility prior to increasing the power level above 485 Mw thermal, and we believe that these changes would make a significant contribution to the safety of operation. These are as follows:
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A.
A third safety injection pump with higher head character-istics than the two existing safety injection pumps would be added.
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i B.
Automatic, low pressure actuation of the third charging pump
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would be provided.
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C.
The reactor would scram automatically on loss of a single
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main coolant pump.
D.
The reactor would scram on low neutron flux.
E.
Automatic rod withdrawal would be eliminated.
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Scram set points would be closer to normal operating conditions. -
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Rod withdrawal at power would be limited to a maximum of I$
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three inches in any eight-hour period.
1 The two safety injection pumps which are presently installed can inject f
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water into the reactor vessel if the primary system pressure falls below l
approximately 260 psig. Upon analysis of the loss-of-coolant type accidents which might occur at the increased power level, Yankee concluded that an additional safety injection pump should be installed which would actuate at a higher pressure. Accordingly, Yankee proposes to add a safety injection pump which will inject water into the primary system if the pressure of the system
._f-falls below approximately 770 psig.
At present, only two of the three charging pumps can be set for automatic operation in the event of low pressure. In order to insure that at least two of the pumps will be available to supply water automatically to the primary
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system in the event of small or medium size pipe breaks, Yankee intends to equip the third pump so that it can be set for automatic actuation on low primary system pressure.
Analysis of the loss of main-coolant-pump type of accident during 540 Mw thermal operation indicated that the loss of the pumping power of a single main coolunt pump could result in a situation where departure from nucleate boiling (DNB) would be approached. Since departure from nucleate boiling could result in fuel element melting, Yankee intends to provide for automatic scram on loss of pumping power of a main coolant pump.
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Based on an analysis of the dropped-control-rod type accident, Yankee concluded that departure from nucleate boiling would not be approached if g
the reactor were scrammed when the power level drops from 540 Mw thermal to 85% of full power. If action were not taken to reduce core reactivity fol-loving the unintentional drop ~of a rod, the power would eventually begin to increase to a new equilibrium level, and damage to the fuel might result from local neutron flux peaking due to the asymmetrical rod pattern present.
Accordingly, any time the reactor is operated at substantial power, the nuclear instrumentation will be set to scram the reactor if there is an un-planned power level reduction corresponding to 15% of full power.
Yankee has also instituted several other procedural and equipment changes to increase the safety of operation. The set points will be closer to normal
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operating conditions (e.g., the over power scram point will be 108% instead of 120% of rated power). Control rod withdrawal to compensate for changes of I
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reactor temperature will be eliminated to preclude the inadvertent addition of an excessive amount of reactivity through this mechanism. When the full operating power level is reached, rod withdrawal will procedurally be limited to 3 inches per eight hour period, which corresponds to a reactivity addition rate of 1.8 x 10-3 in eight hours. Such a restriction would be for-the pur-pose of avoiding the possibility that xenon distribution could result in ex-cessive hot channel factors as a consequence of local neutron flux peaking.
IV.
NEW DNB CORREL.ATION
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In addition to the operational data and experience which Yankee has gained with Core I and the protective features and precautions which would be installed
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or adopted before operation'at the proposed higher power level, another import-e i
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ant consideration is that Yankee is basing its analysis of the effects of the proposed increased power level on Westinghouse's W-2 burnout correlation. A discussion of this new correlation is contained in-a Westinghouse document, l
WCAP-1997, "New DNB (Burnout) Correlations".
The W-2 " correlation" actually consists of two correlations. One is for the quality region and is used for those conditions where burnout results from high enthalpy; the other is used fer operation in the subcooled region where E
burnout results from high local heat flux. Based on the worst hot channel l
factors, as deteomined from in-core instrumentation during Core I operation, i
Yankee reports that the predicted lowest DNB ratio during steady state opera-tion with Core II is ar shown in the following table.
(The DNB ratio is the approximate heat flux at which fuel element melting would be expected to oc-cur divided by the maximum heat flux which will be present in the reactor core.)
The DNB ratios calculated using previous Westinghouse correlations are also given in the table for comparison.
, TABLE I Minimum Steady State Correlation DNB Ratio WAPD-188 2.04 W-1 correlation.
2.17 W-2 correlation 2.20 The W-2 correlation indicates a higher DNB ratio than the other two I
correlations; however, the W-2 correlation represents a "best-fit" of the available data, rather than a lower envelope of most of the available ex-perimental data points. Accordingly, approximately on2 half of the experi-7g
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mentally determined DNB points are at a lower power level than would be predicted by the W-2 correlation.
FL Westinghouse has reported that the W-2 correlation predicts all pub-lished 2000 psi data to within 20%, with a 95% confidence level that the actual DNB ratio is greater than calculated using the W-2 correlation.
Accordingly, the minimum steady state DNB ratio predicted using the W-2 correlation for 540 Mw thermal operation is 1.76 at the 95% confidence level. By expressing the W-2 correlation results in this manner, it is' easier to compare them with the results of correlations used previously.
The thermal characteristics of the Yankee reactor may also be compared with other reactors of the basis of the maximum specific power density in terms of Kw/f t and the maximum heat flux in units of BTU /hr. f t2 Review of the values of these two quantities authorized for other reactors reveals that the maximum specific power and maximum heat flux of the Yankee reactor operating at 540 Mw would be less than those of a number of other power 1
reactors.
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8-V.
ACCIDEhi ANALYSIS Yankee has re-analyzed the various accidents at 540 Mw thermal which were described in its previous application for authorization to ope
- ate at 485 Mv ther=al. This analysis indicates that the only accident which will result in fuel temperatures high enough to damage the fuel is the loss-of-coolant type accident. The analysis showed, further, that even in the loss of coolant accident no melting of fuel rods would take place and that damage to the fuel would be limited to possible distortion of a small number of fuel 3
rods if the safety injection system functions properly. In view of the f act *
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that there is a single high pressure safety injection pump, we do not believe that it is incredible that a loss of coolant accident might occur coincident
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i accident, if the other two safety injection pumps function properly, would probably not result in fuel element melting but may result in the melting te=perature of a portion of the control rod absorber sections being approached l
or exceeded. Since moderator would not be present at the time of such an occurrence, the core should not go critical, and subsequent addition of bor-ated water by the safety injection pumps should not result in criticality being attained. If water which is not suf ficiently borated should enter the core and result in criticality, it is unlikely that the consequences of such an accident would be more severe than the maximum hypothetical accident which Yankee has previously analyzed. This accident involved release of 20% of the gaseous and volatile fission products contained in the core after 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
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. of 485 Mw thermal operation to the vapor container atmosphere. Even in the event of the occurrence of the maximum hypothetical accident following 540 Mw thermal operation, the maximum potential dose rates which would be expected at the site boundary and in the nearby low population zone would be less than the guidelines stated in 10 CFR Part 100. In light of this, we believe that the site is suitable for operation of the reactor at a power level of 540 Mw thermal.
VI.
CHANGES IN TECHNICAL SPECIFICATIONS The proposed increase in maximum operating power would involve c rtain changes to the Technical Specifications attached as Appendix "A" to Yankee's license. Both Yankee and the Division of Licensing and Regulation Staff are presently reviewing the changes which should be made in the Technical Speci-fications to reflect the increased maximum operating power level, if it is authorized.
VII. ADVISORY COMMITfEE ON REACTOR SAFEGUARDS REVIEW Amendment No. 41, dated June 4, 1962, was considered by the AEC's Advisory Committee on Reactor Safeguards (ACRS). In its report to the Comission, dated August 25, 1962, the ACRS stated that it believed that "this reactor may be operated at a power level of 540 Mw thermal without undua hazard to the health and safety of the public."
In the same report to the Com:nission, the ACRS considered a proposal by Yankee Electric Company (Pro}. Jed Change No. 26, dated July 20, 1962) to elim-inste a technical specification which requires power coefficient and moderator L
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. temperature coefficient measurements at 2,000 operating hour intervals.
(The ACRS steted that it believed that "the change in frequency of the 2,000-hour tests may be allowed and that this reactor may be operated at a power level of 540 Mw thermal without undue hazard to the health and safety of the public.") This matter was presented to the ACRS for its information and comment due to the past interest of the ACRS with respect to the matter of potential reactivity coefficient changes due to plutonium buildup in the Yankee reactor core. The Staff believes that the proposed change in the technical specification does not present significant hazards considerations not described or implicit in the hazards summary report and that there is reasonable assurance that the health and safety of the public will not be endangered. Accordingly, the Staff proposes to handle this matter under the Commission's change procedures provided in Section 50.59 of 10 CFR 50.
VII. CONCLUSIONS In view of the foregoing, it is our opinion that the reactor can be operated, as proposed, at a power level of 540 Mw thermal without undue hazard to the health and safety of the public.
Orip.116:d by ?2.:1 H. Brin Robert H. Bryan, Chief Research & Power Keactor Safety Branch Division of Licensing and Regulation A
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