ML20042B414
| ML20042B414 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 03/19/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20042B409 | List: |
| References | |
| NUDOCS 8203250290 | |
| Download: ML20042B414 (12) | |
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s UNITED STATES j
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NUCLEAR REGULATORY COMMISSION 1
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WASHINGTON, D. C. 20555
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THE CONNECTICUT YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-213 HADDAM NECK PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 47 License No'..DPR-61 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Connecticut Yankee Atomic Power Company (the licensee) dated February' '1,1982, complies with the st'andards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commis' ion's rules and s
regulations set forth'in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements l
hcve been satisfied.
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8203250290 920319 DR ADOCK 05000213 PDR
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2.
Accordingly, the license is amended by changes to the Technical 3
Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C(2) of Facility Operating License No.
l DPR-61 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 47, are hereby incor-porated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
[
Dennis M. Crutchfield, C ef i
Operating Reactors Branch #5 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 19, 1982 1
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ATTACHMENT TO LICENSE AMENDMENT NO. 47 FACILITY OPERATING LICENSE NO. DPR-61 DOCKET NO. 50-213 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.
The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.
REMOVE INSERT -
3-15c.
3-15c*
3-16 3-16 3-17 3-17 3-17a f 3-17a 3-17b through 3-17e 3-20 3-20*
- These pages are included merely for correcting editorial errors which occurred during the issuance of Amendment 42, October 8,1981.
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TABLE 3.9-2 (CONTINUED) l I
I ACTION 1 - With the nu'mber of OPERABLE channels less than required by Table 3.9-2, either restore the inoperable channel (s) to l
OPERABLE status within 30 days or be in H0T STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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ACTION 2 - With the subcooling margin monitor IN0PERABLE, determine the subcooling margin once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 3 - With any individual valve position. indicator inoperable, 1
obtain quench tank temperature, level and pressure infor-mation, and monitor. discharge pipe temperature once per shift to determine valve position. Thit action is not
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required if the PORV block valve is closed with power i
removed in accordance with Specification 3.3.C.(6).
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c.
t 3.10 REACTIVITY CONTROL Applicability:
Applies to control rod position during power operation and shutdown margin during suberitical operation except refueling.
Modes 1, 2, 3, 4 and 5.
I Objective:
To define control rod group insertion limits which insure:
(1) an acceptable core power distribution during power operation, (2) a conservative limit on potential reactivity insertion for hypothetical control rod ejection, (3) adequate shutdown margins af ter a reactor trip; that at least 3%21K/K shutdown margin is available durin'g suberitical operation; and (4) that all control rods are within the specified postion with respect to the rest of the bank.
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Specification:
Except for ldw power physics test at or below 10 percent of A.
full power or determination of "just critical" rod positions, operation of the control group banks shall be maintained above the limits shown in Figure 3.10-1.
B.
The maximum worth of any individual control rod in the core at rated power shall not exceed 0.17% AK/K, as measured at the beginning of core life.
C.
The maximum worth of any individual control rod in the l
core with the reactor just critical shall not exceed 0.83%
AK/K, as measured at the beginning of core life.
D.
1)
Except for physics testing, a13% 6K/K shutdown margin shall be maintained during suberitical operation (modes 3, 4 and 5 and the suberitical portion of mode 2).
This shutdown margin may be previded by control rods actually inserted, control rods available to insert and/or soluble boron.
If it is determined that the shutdown margin is less than the above, within 15 minutes initiate and continue boration at 30 GPM of 14,000 PPM boric acid solution or equivalent until the required shutdown margin is restored.
ii) During critical operation (mode 1, and critical portions of mode 2), the control groups shall be limited in physical insertion as shown in Figure 3.10-1.
3-16 Amendment No. 29', 47 w
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.,y With a control group inserted beyond the above-insertion limi~ s except for surveillance testing or rod alignment t
checks either:
(1) Restore the control group to within limits within 2 hou'rs or, (2)
Reduce thermal power within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less taan or equal to that fraction of rated thermal power which is allowed by the group position using Figure 3.10-1 or, a
(3) Be in at least hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
E.
All shutdown bank control rods and control bank control rods shall be operable and positioned within i 24 steps indicated I
position of their group position (as indicated by the RPI) while in Modes 1 and 2.
1.
With more than one rod misaligned (misaligned includes a dropped rod) from its group position by more than i 24 steps indicated position, determine that the shutdopn margin requirements of specification 3.10.D are satisfied within one hour and realign the misaligned rods, or be in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
2.
With more than one rod inoperable, determine that the shutdown requirements of Specification 3.10.D are satisfied within one hour, or be in hot standby -within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
3.
With more than one rod stuck due to being Lamovable as a result of excessive friction or mechanical interference, or known to be untripable, determine that the shutdown margin requirements of Specification D are satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
4.
With one rod stuck, inoperable, or misaligned from its group by more than i 24 steps indicated position:
Restore the stuck or inoperable rod to operable a.
status, or realign the misaligned rod within the above alignment requirements, within one hour, or b.
Power operation may continue up te one full power month, provided that:
(1) within one hour the requirements of Specification 3.10D are satisfied including an allowance for an additional stuck rod; 3-i7 Amendment-No. S9', 47 p
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i (2) The remainder of the rods within the group
'with the inoperable rod are aligned to within i 24 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figure 3.10-1.
The thermal power level shall.
be restricted pursuant to Specification 3.10.D during subsequent operations ;
(3) The shutdown margin requirements of Specification 3.10.D are determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ;
(4) A reevaluation of each FDSA accident analysis is performed within 10 calendar days. This re-evaluation shall confirm that previously analyzed l
results of these accidents remain valid for the duration of operation under these conditions; and (5) A power distribution map is obtained from the moveable incore detectors and the linear heat generation rate is verified to be within the I
limits of Section 3.17 of technical specifications within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
F.
The rod position indication system (RPI) and the rod group position indication system (step counters) shall be operable and capable of determining the rod positions within i 16
, steps while operating in Modes 1 and 2.
1.
With one rod position indicator (RPI) inoperable, (a) Moveient of the rod group which includes the non-indicating rod shall be restricted to i 8 steps from the position last determined prior to loss of the non-indicating rod.
(b)
If the position of the non-indicating rod is not determined indirectly within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, declare the rod inoperable and refer to Specification 3.10.E.4.
l 2.
With one or more rod group position indicators-(step counter) inoperable, either:
(a) Verify that all rod position indicators (RPI) for the affected group are operable and that the most withdraum rod and the least withdrasm rod I
of the group are within a maximum of 32 steps of each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or (b)
Reduce thermal power to less than 75% of rated thermal power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Amendment No. JAf,47 3-17a
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3.
With more than one rod position indicator (RPI), or one rod position indicator (RPI) and its associated rod group position indicator (step counter) inoperable:
Within 8 h'ours, reduce thermal power to less than a.
75% of rated thermal power, and I
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b.
Restore the indicators to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown within i
the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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G.
The rod position indicators (RPI) or the rod group position indicators (step counters) shall be operable and capable of determining the rod position within i 16 steps for each rod not fully inserted while operating in Modes 3, 4, and 5.
With less than the above required rod position indicators operable, immediately open the reactor trip breakers.
H.
The individual rod drop time from fully withdrawn position shall be 12.5 seconds from the fully withdrawn position to the bott6m of the dashpot with Tavg equal to 5300F i 50F and four reactor coolant pumps operating before operating in Modes 1 and 2.
The drop time shall be 1 2.45 see if only 3 reactor coolant pumps are operating while the drop tests are made.
The rod drop time shall be demonstrated prior to crit 3.cality for all control rods following removal of the reactor vessel head or for specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods and at least once per 18 months.
I.
All shutdown control rods shall be fully withdrawn while operating in Modes 1 and 2 except for surveillance testing.
With one shutdown control rod not fully withdrawn, within one hour'either:
(1) Fully withdraw the shutdown control rod or, (2) Declare the shutdown rod inoperable and apply 3.10.E.4.
i Amendment No. 47 l
3-17b s
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3.10 _ REACTIVITY CONTROL (Continued)
BASIS:
The methods of reactivity control to be used are fully explained
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in Section 4.2.2 of the FDSA. The control rod program was developed to insure that three major safety considerations are satisfied throughout core life. They are:
1.
Power distributions (DNB ratios) with equilibrium xenon
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shall be at least as favorable as those used in the safety analysis and shall be within the limits of F and gg Specification 3.17.
2.
Sufficient shutdown margin shall be available to ensure that:
(a) the reactor can be made sufficiently suberitical after a trip from any operating condition, including an allowance for the maximum worth stuck rod, and (b) the reactor does not return to_ criticality for any FDSA Chapter 10 postulated accident.
3.
Potentialejectedrodworthsshallnotexceedthelimits specified in Section 10.2.7 of the FDSA.
As scen in Section 4.3 of the FDSA, the calculated minimum DNB ratio using the rod program is 3.'05 compared with 2.82 using the power distributions considered in the safety analysis.
Since shutdown margin requirements vary throughout core life as a function of fuel dhpletion, RCS boron concentration, and RCS Taverage, the limits of Figure 3.10-1 are set to ensure that the shutdown margin af ter a reactor trip from the most limiting set of reactor conditions meets the design criteria (FDSA Section 4-2-2) of at least 3%6 K/K with all rods inserted and at least 1% A K/K with the highest worth rod stuck out during critical operation.
In addition, operation within the limits of Figure 3.10-1 ensures that sufficient shutdown margin (including a stuck rod) is available to prevent the reactor from returning to criticality during the most limiting FDSA Chapter 10 postulated ac'cident.
Specification 3.10.D(1) insures at least 3%AK/K shutdown margin is available when the reactor is suberitical.
This margin is required to offset the reactivity addition that would occur during a postulated large steamline break accident and boron dilution incident.
l 3-17c Amendment No. JMf[47
f Of thu three safety considerations given above, the third was calculated to be most limiting, and this forms the basis for Specifications A and D.
As shown in FDSA Section 10.2.7, the analysis for the rod ejection was quite conservative and a large margin would exist ;o fuel melting and dispersion.
Power distribution, control rod worths and shutdown margins will be evaluated prior to initial startup and subsequent startups following refueling.
Conformance to the above requirements will be checked at these times and the limit of Figure 3.10-1 adjusted to meet these requirements.
l Specification E limits the time a dropped control rod may be in l
the core because lower fuel depletion and fission product inventory in the vicinity of the dropped rod, relative to the rest of the core, increases the worth of that rod. The lack l
of fuel depletion and lack of fission prtducts other than Xenon
. in the vicinity of a control rod which has been inserted for one full power month will have a negligible effect on the worth of that control rod. Icnon redistribution causes an appreciable increase in the worth of a dropped rod. The increased worth has been calculdted and found to be acceptable.
Should a control rod te dropped, no immediate adverse effects would occur' due to automatic load cut-back as described in FDSA Section 7.2.3.
Operability of the control rod position indicators (RPI) is required to determine control rod positions and thereby ensure complianca with the control rod alignment and insertion limits.
The statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors or a restriction in thermal power; either of these restrictions provide assurance of fuel red integrity during continued operation.
In addition, those aceddent analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
3-17d g-c m
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Misaligt. ment of a single rod of 40 steps has been analyzed and ve,rified for each operating cycle. The 24 steps indicated mis-alignment allowed by the specifications assumes a 16 step mis-alignment allowance, i
The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analysis. Measurement with Tavg equal to 5300F + SOF and with three or four reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
References:
1.
FDSA Section 4.2 2.
TDSA Section 7.2.3 3.
.D.L. Siamann letter to W. G. Counsil, dated November 5, 1979.
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3-17e Amendment No. EI, 47
o Regarding internal' pressure limitations, the containment design pressure o'f 40 psig would not be exceeded if the internal pressure before a major loss-of-coolant accident is maintained in accordance with Technical l
Specification 3.ll.C.
However, 3 psig maximum is sufficient for l
operations of the continuous leakage monitoring system. The containment-is designed to withstand an internal vacuum of.7.5 psig. The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.
The design air recirculation flow rate gith 4 fans operating under saturated conditions of 40 psig and 261 F is 200,000 CFM. The system is designed to perform its function with only 3 of 4 units in operation.
The air filtration system is discussed in detail in FSDA Section 3.6.
The containment spray system in itself can control the contaiiment pressure.
It, therefore, provides a backup to the air recirculation system.
Containment post accident hydrogen venting cac be accomplished by two methods. One uses the containment air particulate monitoring system and the other uses the containment purge exhaust system. These methods are not required in Any short time frame after an accident; it is expected that months may elapse.
In any event the systems used if not operable for maintenance reasons can be readily made operable providing access into the centainment is not required.
Centainment purge is utilized as a backup means of venting hydrogen from' the centainment following'a less-of-cociant accident. The containment tir particulate monitoring system provides the primary means of purging because it provides adequate purge flow to prevent an explosive mixture buildup while allowing fine control of the release of radioactivity bring purges.
k* hen necessary to effect repairs to the containment parge or purge bypass isolation valves, a blank flange must be applied to the 42" purge air exhaust penetration inside the reactor containment so that the containment remains leak tight. This renders the purge system inoperable for a finite time.
Seven days is considered a reasonable length of time for repair parts to be received, installed and the system retested for leak tightness and returned to service.
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(Correction) 3-20 Amendment No. J7,.6, 47
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