ML20042B415
| ML20042B415 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 03/19/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20042B409 | List: |
| References | |
| NUDOCS 8203250291 | |
| Download: ML20042B415 (4) | |
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UNITED STATES f'
NUCLEAR REGULATORY COMMISSION 5
tj WASHINGTON, D.C 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 47 TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET NO. 50-213
1.0 BACKGROUND
The staff review of the LER's and Technical Specification (TS) requirements related to the Control Rod Position Indication Systems (RPI) at Westinghouse PWRs determined that a wide variation exists in the number of LER's received and the TS requirements.
2.0 DISCUSSION Westinghouse has performed safety analyses for control rod misalignment up to 15 inches or 24 steps (one step equals 5/8 inch). Since analysis of misalignments in excess of this amount have not been submitted, we have imposed an LC0 restricting continued operation with a misalignment in excess of 15 inches. Because the analog control rod position. indication system has an uncertainty of 7.5 inches (12 steps), when an indicated deviation of 12 steps exists, the actual misalignment may be 15 inches.
This is because one of the coils, spaced at 3.75 inches, may be failed without the operator's knowledge. The Standard Technical Specifications (STS) were written to eliminate any confusion about this, and restrict deviations to 12 tr.dicated steps. Surveillance requirements, on the indication accuracy of 12 steps, were also prepared to ensure that the 15 inch LC0 is met. 'Since there is no difference intended in requirements issued for any Westinghouse reactor, plants with Technical Specifications written in different terms of misalignment should consider the 12 step instrument inaccuracy when monitoring rod position.
A related problem is that the installed analog control rod position indicating system equipment may act, is some areas, be adequate to maintain the control rod misalignment specification requirement because of drift problems in the calibration curves. This is evidenced by numerous LER's concerning rod l
position indication accuracy.
In these cases, the uncertainty may be more than 12 steps.
t Connecticut Yankee Atomic Power Company (CYAPCO) was requested by letter dated November 5,1979, to review the Technical Specifications for the Haddam Neck Plant to ensure that the control rods are required to be main-tained with + 12 steps indicated position and that the rod position indication system is accurate to within + 12 steps.
8203250291 820319 PDR ADOCK 05000213 l
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By letter dated January 30, 1980, CYAPCO responded to our request and stated that TSs would be provided but, since the Haddam Neck installed-system is different than the model system used in the STS, modifications to the TS would be required.
In a May 24, 1981 letter, CYAPCO informed the NRC that their study of the problem had resulted in the following conclusions.
1.
The rod position uncertainty due to instrumentation inaccuracy has been determined by statistical analysis to be + 16 steps (9 3/8" per step).
2.
For single control rod misalignments up to + 32: steps ~(in-strument inaccuracy plus allowable misaligniient of 16 steps before alarm), the peaking factor increase associated with the misaligned rod does not' invalidate the design or licensing bases for Haddam Neck.
3.
As calibration of the RPI system is accomplished at hot zero power condition's (535'F), the indicated rod positions are uniformly shifted as reactor coolant temperature rises to hot full power conditions (555*F).
CTAPC0 further concluded that the present system provides safe operation and functions efficiently and effectively.
Since there is an effort to convert the Haddam Neck TS to the STS, CYAPCO proposed defert'ing changes to incorporate our request subject to that conversion. Subsequent conver-sations resulted in an agreement to revise Section 3.10 " Reactivity Control"'
of the TS to incorporate the staff request on control rod positioning and update and clarify the other requirements. By letter dated February 1,1982, CYAPCO proposed this revision.
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3.0 EVALUATION We have reviewed the February 1,1982 CYAPC0 application and compared the proposed changes'against the existing Haddam Neck TS requirements and present staff position.
Since the control rod drive mechanisms move the rod only 3/8 inch per step instead of the 5/8 inch per step assumed in the STS, adjustments were calcu-lated to ensure that our positions were met. For example, the STS require-ments are based on 15 inches; this corresponds to 40 steps for the Haddam Neck system. Similar correlations were performed on the rod position '
indication system and the proposed requirement to maintain the control rods with + 24 steps indicated position provides the requirement requested in our NovemFer 5,1979 request. Therefore, we find the proposed changes to Sections E and F of Specification 3.1.0 to be acceptable.
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3-Additional changes and additions were proposed to Specification 3.10 to clarify existing requirements and to add additional requirements.
We have reviewed these changes and have found the following:
1.
Sections A, B and C restate existing Sections A, D and E.
2.
Section D restates the requirements presented in the existing Section'F and clarifies the requirement.
3.
Section H adds a new requirement for Control Rod drop time which is consistent with the assumed drop time used in the accident analyses and is acceptable.
4 Section I adds a new requirement that the Shutdown Control-Rods be fully withdrawn during power operations. This addition is consistent withNRCguidangeandisacceptable.
Since these clarifications and additions provide for continued safe operation of the Haddam Neck Plant, we conclude that'they are acceptable.
4.0 ENVIRONMENTAL CONSIDERATION
We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2).there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission'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.
6.0 ACKNOWLEDGEMENTS The following NRC personnel have contributed to this evaluation:
P. Wagner C. Tropf Date:
March 19, 1982 l
.. t Evaluation - By letters dated December 4,1979, and July 22, 1981, the licensee stated that a normally open motor-operated isolation valve will be installed in the common corssconnect line between the tow turbine driven AFM pumps.
Manual control of this valve will be provided in the control room.
This valve will be capable of local manual operation if required. The intended modification will permit isolation of the two AFW train in the event of a break in the common c'rossconnect line.
By letter dated December 4, 1979, the licensee stated that the third motor driven pump will onl1y be aligned temporarily using a pipe spool piece during startups to supply feedwater to the steam generators.
By letter dated July 1, 1981, the staff requested that the licensee upgrade this pump to meet the power diversity position as contained in Branch Technical Position ASB 10-1.
By letter dated August 27, 1981, the licensee provided additional information. We will report resolution of this matter in a supplement to this SER.
2.4.4 Considerations Based on the Systematic Evaluation Procram 1.
The AFW system itself islot des'igned to withstand a passive failure'at
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all points within the system. A pipe break in a nomally pressurized portion of the AFW system can be isolated by operation of manual valves outside the control room. An alternate flow path to all four S/G's would be available following such isolation. The motor driven main feedwater pumps may also be available in this event since no transient should result to cause a loss of non-vital power.
For the same reasons, the main feed pumps may also be available following a break in any portion of the AFW system that ir not nomally pressurized even though the AFW system (CORRECTION - Safety Evaluation supporting Amendment No. 44 to license dated ;1-20'51)
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