ML20141F814
| ML20141F814 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 04/14/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20141F467 | List: |
| References | |
| NUDOCS 8604230179 | |
| Download: ML20141F814 (17) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION a
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDr!ENT NO. 74 TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HA00AM NECK PLANT DOCKET NO. 50-213
1.0 INTRODUCTION
By letter dated December 11, 1985, the Connecticut Yankee Atomic Power Company (CYAPCO) submitted a request for changes to the Haddam Neck Plant technical specifications. By letter dated February 18, 1986, CYAPCo submitted the reload technical report in support of cycle 14 operation.
The amendment changes technical specifications that are directly related to the fuel cycle design and safety analyses for plant operation during cycle 14. The technical specification changes include: 1) the definition of quadrant power tilt ratfo; 2) setpoints for protection instrumentation;
- 3) isothermal coefficient of reactivity; 4) limiting heat generation rates;
- 5) power distribution monitoring and controls; and 6) reactor coolant system flow, temperature and pressure.
On February 26, 1986, CYAPCo experienced a dropped fuel assembly causing damage to three (3) fuel assemblies in the core. As a result, CYAPCo revised and reanalyzed the fuel loading pattern for cycle 14 The revised analyses were submitted by letter dated April 3, 1986. This evaluation is based upon information from both the February 18 and April 3,1986 technical reports. The CYAPCo reload reanalyses did not result in any revisions to the technical specification requested in the December 11, 1985 application.
8604230179 860414 DR ADOCK 05 23
s A Notice of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing rela'ed to the requested action was published in the Federal Register on January 29, 1986 (51 FR 1713). No comments or requests for hearing were received.
2.0 EVALUATION The Haddam Neck cycle 14 was evaluated for a nominal cycle life of 356 Effective Full Power Days (EFPD) with a mregin of 8 EFPD (raximum cycle life of 364 EFPD). The Haddi a N,eck core consists of 157, 15 x 15 assembifes containing 204 fuel rods, 20 control guide tube and one in-core instrument guide tube. All fuel assemblies, except four, have stainless steel clad fuel pins (SS304) 16.5 mil thickness. Four assemblies (batch ISB) have zircaloy cladding of 27.0 mil thickness.
They have been designated as lead test assemblies (LTA). The new fuel is batch 16 (52 assemblies) and it has bevelled disk and uranium dioxide cylindrical pellets of minimum fuel density of 94.9% of theoretical.
The batch 16 average nominal fuel assembly loading is 411.5 kg.
In addition to batch 16, a twice burned assembly discharged at the end of cycle 8, will be inserted in the central location (J36) of cycle 14 and four batch 11 assemblies, discharged at the end of cycle 11, will be I
reinserted in locations synnetric to location Hl..The 52 fresh assemblies have 4% enrichment and will be placed in peripheral locations of the core in a high leakage out-in configuration.
2.1 Fuel System Design Evaluation 2.1.1 Fuel Assembly Hechanical Design Cycle 14 contains fuel assemblies from batches 9, 11, 14, 15A, 158 and 16. All fuel assemblies are mechanically interchangeable. The upper and lower spacer grids for the 4 LTAs are slightly moved to accommodate the shorter LTAs which also have lower enrichment (3.41%)
and thicker cladding to be compatible with the rest of the stainless steel fuel assemblies.
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2.1.2 Cladding Stress Strain and Collapse The batch 11 assemblies in cycle 14 are the most limiting (highest in-core exposure) with respect to clad creep collapse. However, analysis estimates indicate that the collapse time is longer than the expected residence time in cycle 14. All cycle 14 fuel rods have been analyzed for stress and it was estimated that all stresses were about 13.8% below the required minimum of 2/3 of the unirradiated yield strength. For the cladding strain estimate, the worst case of heat generation rate was combined with the upper tolerance value of the clad. With these conservative assumptions, the strain was estimated to be less than the required 1%. The cumulative fatigue usage factor for the stainless steel 304 clad fuel and for, the zircaloy-clad fuel was estimated to be 0.2 and 0.4 respectively. Both are much lower than the ASME code allowed value of 0.9.
Based on the above results, the cladding design was found to be acceptable.
2.1.3 Fuel Thermal Design All cycle 14 fuel is thermally similar and, in particular, batch 16 is identical to batch 15A. The batch 158 assemblies (the 4 LTAs) are less limiting than any of the stainless steel clad assemblies.
The fuel thermal design calculations have been performed with the TACO 2 code and conservative values of the pertinent parameters were used which is acceptable.
2.2 Nuclear Design Evaluation The nuclear design parameters for cycle 14 were generated using PDQ07 and were compared to the corresponding quantities for e T1e 13.
Allowing for the fact that the BOL core average burn-up
.2 lower for cycle 14, the critical boron concentrations are respectively higher.
The maximum stuck rod worth is used in the shutdown margin calculation.
The ejected rod worths were also analyzed for 40 EFPD beyond the end of core life. Other cycle-14 nuclear design parameters i.e.
Doppler and moderator coefficients, zenon and boron worths and the effective delayed neutron fraction compare to those of the reference cycle.
The cycle 14 nuclear design meets applicable acceptance criteria, and was evaluated using conservative parameter values and an acceptable methodology. Therefore, we find it acceptable.
2.3 Thermal Hydraulic Design Evaluation The new fuel assemblies (batch 16) have been fabricated to the same tolerances as batch 15A for the reference cycle 13. Batch 158 consisting of the 4 zircaloy-clad LTAs as the remainder of the core have been evaluated using the TACO 2 program. The LTAs are no more limiting than the stainless steel assemblies and the sr.fety limits calculated for cycle 13 are also applicable to cycle 14.
The design data for the cycle 14 analysis are conservative in that they are based on 9% core bypass flow,100% power level and a power distribution which yields the most limiting power peak throughout the cycle (assembly G-2).
The hot pin power has been increased by the required values of uncertainties and penalties. The resulting minimum departure from nucleate boiling ratio (MDNBR) is 2.50 compared to the design value of 1.30.
TAC 02 also estimated the maximum steady-state fuel temperature to be 4300*F, which is less than the design value of 4780 F.
Therefore, the thermal hydraulic design was estimated using conservative assumptions, approved methodology and yielded acceptable results. Hence, the thermal-hydraulic design is acceptable.
(No fuel rod bow penalty has been estimated due to the stainless steel cladding.
For the 4 LTAs it is not required).
2+
Accident and Transient Analysis Evaluation The expected operating conditions for cycle 14 include lower temperature and pressure, however, the thermal margin for steady-state operation and the accident analysis is equivalent to that of the original analysis. Several of the accidents have been re-analyzed -
after the first cycle: LOCA was re-analyzed during cycle 2 in compliance with the interim acceptance criteria; the rod drop accident was re-analyzed during cycle 5 due to modification of the rod bank configuration and the steam line break accident was re-analyzed during cycle 8 to support a change in the sequencing of the charging pumps. Fuel densification, gap increase and repressurization modifications were incorporated in cycle 12 (batch 14). The results of the cycle 14 accident analysis are compared to each of the currently valid analyses. Particularly for fuel densification, the batch 16 fuel density is higher than that of batch 14, hence, the conclusions of that analysis are still valid (Reference 6).
2.4.1 Control Rod Withdrawal For the uncontrolled rod withdrawal, we distinguish two cases i.e.,
from subcritical condition and from power. Assuming that no other protection trip will function except for the overpower trip, the most conservative case is the one assuming the fastest reactivity insertion.
These assumptions for the controlling parameters for this transient assure a conservative analysis.
In the original analysis, the least negative value of the Doppler coefficient was -1 pcm/*F. the most positive value of the moderator temperature coefficient was +10 pcm/'F and the maximum differential control rod group worth was 90 pcm/ inch.
(percentmil pcm = 10-5 k/k) However, the corresponding values for cycle 14 will be -1.72 pcm/ F (Doppler coefficient), -3.93 pcm/*F (moderator temperature coefficient) and 39.84 pcm/ inch (differential 1
rod worth), therefore, the original analysis conservatively bounds the cycle 14 results.
The uncontrolled rod withdrawal from power could be terminated by the overpower trip or the variable low-pressure trip. Whether one or the other would be activated depends on the rate of reactivity insertion, for example, a very fast rate will be terminated by the overpower trip. A parametric survey of the reactivity insertion rates indicates that the most severe trnnsient corresponds to the
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case when the overpower and the variable low-pressure trips are activated simultaneously. However, the margin for the operating pressure and temperature with respect to the reactivity insertion transient was reassessed for cycle 7 (Reference 7) and was found i
to be no less than the original analysis. This conclusion is also applicable for cycle 14. Therefore, the rod withdrawal transients (from zero and full power) have been conservatively analyzed and shown to have the same or greater margin than the originally approved analysis and are acceptable.
l 2.4.2 Isolated Loop Startup Incident l
If during operation an isolated loop is brought into operation, an increase in core reactivity (and power) can result due to boron concentration and temperature mismatch. To prevent this from happening, a temperature interlock valve prevents the cold leg stop valve from opening if the temperature difference is in excess of 20*F from the same location in the hottest operating loop.
In addition, operating procedures have been established to prevent isolated mismatch. Nevertheless, an analysis has been performed to evaluate the potential transient. The initial analysis assumed a moderator temperature coefficient of -35 pcm/ F and a Doppler coefficient of -0.5 pcm/ F.
The corresponding values for cycle 14 are -27.16 pcm/*F and -1.40 pcm/"F respectively.
Therefore, the isolated loop mismatch transient is very conservatively bounded by the initial analysis and is acceptable.
2.4.3 Boron Dilution Transient A boron dilution transient can result in reactivity increase with subsequent power increase or loss of shut down margin.
For cycle 14 with a 80L shutdown margin of 3560 pcm and a maximum flow rate of 180 gpm, the maximum reactivity addition rate at power is 0.90 pcm/sec..
Hence, the required time to lose shutdown margin is over 55 min, which is considered adequate for the operator to be alerted and take the appropriate action to halt the dilution process. This time interval is acceptable.
2.4.4 Excess Feedwater Transient The excess feedwater transient is a result of oversupply of feedwater flow to one or more steam generators. The excess water flow will lower the cold leg temperature, increase the reactivity and core power.
In the initial analysis, the moderator temperature and Doppler coefficients were -35.0 pcm/ F and -0.5 pcm/*F, respectively. However, for cycle 14 the estimated values are -27.16 pcm/*F and -1.40 pcm/ F, respectively.
l These values conservatively bound.those assumed in the initial I
analysis with respect to reactivity feedback. Therefore, the results of this analysis are acceptable.
2.4.5 Excessive Load Increase Load increase (beyond its normal limits) will result in an increase of steam flow, a decrease of the reactor coolant temperature and an increase in reactor power.
In the initial analysis, the moderator temperature and Doppler coefficients were -35.0 pcm/ F and -0.5 pcm/*F respectively. The calculated values for cycle 14 are -27.16 pcm/*F and -1.40 pcm/*F respectively. Therefore, the cycle 14 load increase transient is conservatively bounded by the results of the initial analysis and are acceptable.
2.4.6 Control Rod Drop Transient This transient is the result of a sudden release and drop of a control rod into the core. This transient was analyzed in cycle 5 for modifications of the control rod bank B (Reference 8).
For cycle 14 the computed values of the moderator temperature coefficient, the Doppler coefficient and the maximum dropped rod worth are all bounded by the referenced analysis. For cycle 14 the NDNBR is 1.45 and the maximum centerline temperature is 4411"F. The 1.45 value is higher l
than the design value of 1.30 and the temperature of 4411 F is below the 4700 F melting temperature of the fuel. These results are acceptable.
2.4.7 Control Rod Ejection Transient A control rod ejection transient results when a control rod is inadvertently rapidly ejected from the core. This represents the most rapid reactivity insertion that can reasonably be postulated.
The controlling cycle 14 BOL parameters are all conservatively bounded by the corresponding values of the initial analysis, except for power distribution which yields a hot channel factor F =7.94 compared to q
F =5.00 for the initial analysis. However, the cycle 14 analysis was q
performed with 830 pcm rod worth that is considerably higher than 301 pcm, which is the estimated rod worth for cycle 14. The difference in F was taken-into account by increasing the peak fuel centerline q
temperature to account for the higher F value (Reference 9). With q
the very conservative rod worth, the peak rod temperature is about 4300*F, i.e., less than the fuel melting point of 4700 F.
2.4.8 Loss-of-Coolant Accident The loss-of-coolant accident has been addressed for the type of fuel present in cycle 14 The resultant limiting linear heat generation rates are included in the axial offset curves.
2.4.9 Loss-of-Flow Transient i
The most severe transient occurs with the reactor at full power coircident with a loss of all four coolant pumps. The coolant temperature will increase rapidly which can result in DNB and fuel 4
damage.
In the initial analysis, the Doppler and the moderator temperature coefficients were -0.50 pcm/ F and 10.00 pcm/*F while for cycle 14 the values are -1.40 pcm/*F and -8.21 pcm/*F, respectively.
The cycle 14 values are conservatively bounded by the initial analysis.
Therefore, these results are still valid.
2.4.10 Steam Line Break For the Haddam fleck design, it has been estimated that the most limiting l
steam line break is a double ended rupture of a '24 inch steam line, j
upstream from the steam line isolation valves. Rapid loss of steam will cause a reduction of the primary coolant temperature that will insert reactivity and increase reactor power. The overpower trip will terminate the transient.
For the initial analysis, the moderator l
coefficient was -35.0 pcm/*F and the Doppler coefficient was -0.50 pcm/*F.
This combination of parameters results in peak power before reactor trip and, therefore, before DNBR.
For cycle 14, the moderator and Doppler coefficient are -27.16 pcm/ F and -1.40 pcm/*F, respectively, which are conservatively bounded by the initial analysis. The maximum loss of shutdown margin has been re-evaluated to support plant modifications for the automatic initiation of auxiliary feedwater. Cycle 14 was re-evaluated and adequate shutdown margin l
is provided. The results of the steam line break transient are acceptable.
2.4.11 Steam Generator Tube Rupture Steam generator tube rupture is important as far as the radiological consequences are concerned. However, the accident is independent of core loading and remains valid and acceptable.
2.4.12 Loss-of-load Transient In the loss-of-load transient, we assume rapid reduction of the turbine-generator load. A turbine overspeed will cause a reactor trip, but the coolant temperature and pressure will rise.
The severity of the transient is determined by the temperature rise that in turn is determined by the moderator temperature coefficient and i
the Doppler coefficient. The most' adverse results are obtained with the most positive moderator coefficient and the least negative Doppler coefficient.
In the original analysis, the corresponding values were
+10.00 pcm/*F and -0.50 pcm/*F, which conservatively bound the,
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corresponding values for cycle 14, f.e., -8.21 pcm/ F and -1.40 pcm/*F, respectively. Therefore, the results of the initial analysis are still valid and acceptable.
2.4.13 Loss-of-Feedwater Transient i
The loss-of-feedwater transient assumes a reduction in feedwater to the steam generators without corresponding reduction to the steam supply, thus, reducing the water inventory in the secondary system.
The primary coolant temperature and pressure will rise and will generator a reactor trip. Pressure vessel protection is provided by operation of the pressure relief valves. The minimum DNBR will be reached if there is a total loss of feedwater combined with the maximum positive value of the moderator temperature coefficient and i
the minimum negative Doppler coefficient.
The values for the initial analysis were 10.00 pcm/*F and -0.50 pcm/*F, respectively.
For cycle 14, the corresponding values are -8.21 pcm/*F and -1.40 pcm/*F, which are conservatively bounded by the original analysis for the purposes of the feedwater transient. Therefore, the initial results are still valid and acceptable.
2.4.14 Fuel Handling Accident This accident which considers the possibility of a fuel mishandling operation is independent of core loading.
2.4.15 Waste Gas Release This accident considers the unintended and uncontrolled release of the zenon and krypton fission gases from the decay tank.
This l
accident is independent of Core loading.
i 2.4.16 Hypothetical Accident This accident is based on the assumption of a gross release of fission products in the containment. The consequences of such release are independent of core loading. l
2.4.17 Summary We have reviewed the information presented for the accident and transient analyses and concluded that the results were obtained using approved methods, meet the criteria applicable for the Haddam Neck reactor and are acceptable.
2.5 Evaluation of Technical Specification Changes 2.5.1 Introduction It is proposed by the licensee to modify the following technical specifications:
1.18 Definitions 2.4 Maximum Safety Setting Protective Instrumentation 3.16 Isothermal Coefficient of Reactivity 3.17 Limiting Linear Heat Generation Rate 3.18 Power Distribution Monitoring and Control, and 3.20 Reactor Coolant System Flow, Temperature and Pressure.
The above technical specifications are directly related to the fuel cycle design and the LOCA safety analysis. We shall examine the acceptability of each technical specification.
L 2.5.2 Definitions (Technical Specification 1.18)
The Quadrant Power Tilt Ratio will no longer be based on core delta-T, but will use the excore neutron detectors to determine tilt and verify compliance.
(Core delta-T will be used only as a backup).
i This change is consistent with cycle 14 plant modifications and the j
Westinghouse Standard Technical Specifications and is acceptable.
2.5.3 Maximum Safety Setting Protective Instrumentation (Technical Specification 2.4)
There are three changes in this specification:
the variable low pressure trip setpoint, nuclear setpoint related to the quadrant power tilt and the three loop nuclear overpower setpoint.
The variable low pressure trip setting change was required by the relocation of RTDs from the rector coolant pump suction piping to the pump discharge piping.
In the new location there are no streaming effects to be considered. The trip setting has been changed to reflect a delta-T=45'F at full power instead of delta-T=40*F. This change is more accurate because it avoids the streaming inaccuracies and, therefore, is acceptable.
j The change related to the quadrant tilt power was an editorial footnote modification for consistency with the new quadrant tilt power definition and to remove an implied quadrant tilt power limit of 10%. This change is acceptable.
The third change reduced the three pump operation overpower trip setpoint from 84% to 74% of the rated power.
This change brings this specification into consistency with the requirements of specification 3.3 which establishes the upper limit of three pump operation at 65% of rated thermal power. Taking into account a 9% margin the equivalent level of protection is 74%.
This change is conservative and acceptable.
l 2.5.4 Isothermal Coefficient of Reactivity (Technical Specification 3.16) i l
l The proposed change narrows the range of allowable temperature I
coefficients and is based on a LOCA analysis consistent with the
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interim criteria for emergency core cooling systems.
Cycle specific parameters (to be derived from the startup physics test program) 1 will be used to convert the isothermal coefficient to a moderator temperature coefficient. This change adopts a more conservative isothermal coefficient range and is acceptable.,
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j 2.5.5 Limiting Linear Heat Generation Rate (Technical Specification 3.17)
The revised values of the Linear Heat Generation Rate (LHGR) were required by steam generator tube plugging and sleeving, by a reduction in the engineering factor and by a large break LOCA re-analysis. The resulting LHGR limits assure that the peak clad temperature does not exceed the 2,300*F which is the accepted limit for the Haddam Neck plant. The 4 LTAs which are Zircaloy-4
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cladded are not limiting in this calculation. The reduced engineering factor has been reviewed and approved elsewhere in this safety evaluation. Therefore, the new values of the LHGR result in peak clad temperatures, which are within approved limits for this plant, and, hence, are acceptable.
2.5.6 power Distribution Monitoring and Control (Technical Specification 3.18)
The axial offset limits are established to prevent exceeding the LHGR at the maximum allowable rod insertion, therefore, the requirement for margin allowance for rod insertion is not necessary. The Cycle 14 design includes a conservative evaluation of rod design so that no achievable positive offset would cause the LHGR limits to be exceeded. The proposed generic axial offset limits have been made more restrictive, hence, the propcsed bank insertion of 280 steps is acceptable. The axial offset limits have been expanded to include (explicitly) three loop operation (Fig. 3.18 Reference 1) and revised four loop limits.
Because three loop offset limits have been added, the power for three loop operation is no longer required and has been eliminated.
The proposed changes in this specification are, therefore.
acceptable.
2.5.7 Reactor Coolant Systen Flow, Temperature and Pressure (Technical Specification 3.20)
This specification change was required by the anticipated steam generator tube plugging and sleeving and a re-evaluation of the core bypass flow. The revised bypass value has been reviewed and approved elsewhere in this evaluation. Although the net core flow has been reduced, it remains above that required for the transient
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analysis and safety limit curves given in technical specification 2.1 and the proposed change is acceptable.
l 2.6 Evaluation of Changes in the Core Bypass Flow and the l
Engineering Hot Channel Factor l
As discussed above, the core bypass flow rate and the engineering l
hot channel factors have been changed and used for the cycle 14 reload calculations. Our evaluation of these changes (references 5 and 10) follows.
2.6.1 Core Bypass Flow The core bypass flow is defined as the total amount of the reactor l
coolant flow which bypasses the core region.. Paths for this bypass baffle-barrel region, the head cooling spray nozzles, the fuel are:
assembly baffle plate cavity, the outlet nozzle and the thimble tubes.
In the early design of the Haddam Neck Plant, a cruciform i
control rod was assumed in the calculation of the core bypass flow.
However, the plant has actually been using the rod cluster control assemblies which allow much smaller bypass flow. Nevertheless, the calculational basis always assumed the conservative 9% bypass flow.
Recent extensive generator tube plugging required a reconsideration of the bypass flow. The plant specific data were supplied by CYAPC0 (Reference 5) to Westinghouse who performed the analysis, in the same manner as for other Westinghouse plants. Conservative assumptions were used such as fully withdrawn control rods, isothermal flow and a withdrawn instrumentation flux detector. A conservative I
1 1
1 uncertainty was assumed for the instrumentation tube and thimble tube bypass. The data, the calculational method and the assumptions, support the result of the requested 4.5% core bypass flow, which we find acceptable.
2.6.2 Engineering Hot Channel Factor The early design of the fuel pellets and fuel rods was subject to manufacturing and engineering tolerances which yielded an engineering E
hot channel factor Fq =1.04.
However, recently the fabrication tolerances have improved. Manufacturing tolerance data for the fuel assemblies currently used in Haddam Neck were used to detennine F E g,
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Assuming 95% confidence level for 95% of the sample population and using the methodology describeo in Reference 11, the F E value is q
1.02.
The methodology has been used previously for the Midland FSAR which was approved by the staff and likewise for the Oconee E
plants. Based on the above, we find that Fg =1.02 is acceptable.
2.7 Summary l
We have reviewed the information submitted by CYAPC0 to support the Haddam Neck cycle 14 reload. The submittals included a reload report, request for Technical Specification changes, a re-evaluation of the core bypass flow and a re-evaluation of the hot channel engineering factor. The revised values of the core bypass and the s
engineering factor of the 4.5% and 1.02, respectively, were found to be acceptable.
In the reload report, the fuel system design, the fuel nuclear design, and, the core thennal hydraulic design were performed based on conservative assumptions and approved methodologies and were found to be acceptable. The accident and transient analysis considered the entire spectrum of transients and accidents of the original submittal which were approved by the staff.
In addition, the applicant, in a meeting with t' e staff (Reference 1), agreed to h
provide a fuel misloading analysis before the end of cycle 14.,
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l The reload of cycle 14 of the Haddam Neck Plant does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety, and therefore, does not involve a significant hazards consideration.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation The Commission has previously issued a proposed finding that exposure.
this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Consnission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
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5.0
References:
i 1.
Letter, J. F. Opeka, CYAPC0 to Director NRR, dated December 11, 1985.
i 2.
Memorandum from C. Grimes, Director, Integrated Safety Assessment to J. Knight, Assistant Director for PWR-A, dated December 20, 1985.
3.
BAW-1878, " Connecticut Yankee Atomic Power Company, Haddam Neck Plant, Technical Report Supporting Cycle-14 Operation," dated December 1985.
4 Letter, J. F. Opeka, CYAPC0 to C. Grimes NRR, "Haddam Neck Plant.
Additional Information for Proposed Revision to Technical Specifications Cycle 14 Reload," dated January 9,1986.
5.
WCAP-11042, " Northeast Utilities Service Company Haddam Neck Plant, Core Bypass Flow Summary Report," Westinghouse Electric Corporation, September 1985.
6.
WCAP-8213, " Effects of Fuel Densification on the Connecticut Yankee Reactor," Westinghouse Electric Corporation, dated October 1973.
7.
"Haddam Neck, Cycle VIII Operation and Proposed License Amendments,"
Connecticut Yankee Atomic Power Company, dated May 3, 1976, Supplement June 27, 1976.
8.
YAEC-1080, " Connecticut Yankee Cycle 5 Modification of Control Rod Bank B," CYAPC0, dated December 10, 1974.
9.
WCAP-8213. " Effects of Fuel Densification on the Connecticut Yankee Reactor," Westinghouse Electric Corporation, dated October 1973.
10.
F. Akstulewicz, Project Manager, ISAP DPWR-8, " Meeting Summary concerning the Review of the Cycle 14 Reload Application,"
dated March 3, 1986
- 11. EAW-1912. " Connecticut Yankee Atomic Power Company, Haddam Neck Plant, Technical Report Supporting Cycle 14 Operation Redesign,"
dated March 1986.
6.0 ACKNOWLEDGEMENT
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This Safety Evaluation has been prepared by L. Lois, RSB, DPWRL-A.
Dated: April 14, 1986 '
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