ML20209E487
| ML20209E487 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/27/1986 |
| From: | Colburn T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20209E490 | List: |
| References | |
| TAC-52665, TAC-52666, NUDOCS 8609100489 | |
| Download: ML20209E487 (18) | |
Text
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UNITED STATES
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i NUCLEAR REGULATORY COMMISSION U
': j WASHINGTON, D. C. 20555
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WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.104 License No. DPR-24 l.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated October 25, 1983 as revised February 7 and April 18, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and 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 Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied, i
1 8609100489 860827 PDR ADOCK 05000266 P
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.
DPR-24 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.104, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMISSION f
b Timothy G. Colburn, Project Manager PWR Project Directorate #1 Division of PWR Licensing-A
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 27, 1986.
'n a merg'o UNITED STATES
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8 NUCLEAR REGULATORY COMMISSION o
E WASHINGTON, D. C. 20555
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WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.107 License No. DPR-27 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated October 25, 1983 as revised February 7 and April 18, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and 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 Comission; C.
There is reasonable assurance (i) 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the conmon 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 Comission's regulations and all applicable requirements have been satisfied.
2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.
DPR-27 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.107, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f
/t Timothy G.
olburn, Project Manager PWR Project Directorate #1 Division of PWR Licensing-A
Attachment:
Changes to the Technical Specifications Date of Issuance: August 27, 1986
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ATTACHMENT TO LICENSE AMENDMENT NOS. 104 AND 107 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 15.3.6-2 15.3.6-2 15.3.6-3 15.3.6-3*
15.4.4-1 15.4.4-1 15.4.4-2 15.4.4-2 15.4.4-2a 15.4.4-2b 15.4.4-5 15.4.4-5 15.4.4-6 15.4.4-6 15.4.4-6a 15.4.4-6a 15.4.4-12 15.4.4-12 15.4.4-13 15.4.4-13 15.4.4-16 15.4.4-16
- Separate pages for Unit 1 and Unit 2.
j
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C.
Containment Purge Supply and Exhaust Valves The containment purge supply and exhaust valves shall be locked closed and may not be opened unless the reactor is in the cold shutdown or refueling shutdown condition.
a.
One of the redundant valves in the purge supply and exhaust lines may be opened to perform the repairs required to conform with TS 15.4.4.II.B.
The time duration and shutdown require-ments of TS 15.4.4.II.B.1.b shall be applied.
D.
Containment Structural Integrity The structural integrity of the reactor containment shall be maintained in accordance with the surveillance criteria specified in 15.4.4.V and 15.4.4.VII.
1.
If more than one tendon is observed with a prestressing force between the predicted lower limit (PLL) and 90% of the PLL or if one tendon is observed with prestressing force less than 90% of the PLL, the tendon (s) shall be restored to the required level of integrity within 15 days or the reactor shall be in hot standby within the next six hours and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. An engineering evaluation of the situation shall be conducted and a special report submitted in accordance with specification 15.4.4.VII.D within 30 days.
2.
With an abnormal degradation of the containment structural integrity in excess of that specified in 15.3.6.D.1, and at a level below the acceptance criteria of specification 15.4.4.VII, restore the containment structural integrity to the required level within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within the next six hours and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Perform an engineering evaluation of the containment structural integrity and provide a special report in accordance with specification 15.4.4.VII.D within 30 days.
Basis The Reactor Coolant System conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the Reactor Coolant System ruptures.
15.3.6-2 Unit 1 - Amendment No. H M.104 Unit 2 - Amendment No. M, M,107
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..--J-The shutdown conditions of the reactor are selected based on the type of activities that are being carried out. When the reactor head is not to be removed, the specified cold shutdown margin of 1% ak/k precludes criticality under any occurrence. During refueling the reactor is subcritical by 5%
ak/k. Positive reactivity changes for the purpose of rod assemb*1y testing will not result in criticality because no control bank worth exceeds.3%.
Positive reactivity changes by boron dilution may be required or small concentration fluctuations may occur during preparation for, recovery from, or during refueling but maintaining the boron concentration greater than 1800 ppm precludes criticality under these circumstances.
1800 ppm is a nominal value that ensures 5% shutdown for typical reload cores. Should continuous dilution occur, the time intervals for this incident are discussed in Section 14.1.5 of the FSAR.
i Regarding internal pressure limitations, the containment design pressure of I
60 psig would not be exceeded if the internal of-coolant accident were as much as 6 psig.II) pressure before a maj The containment is designed to withstand an internal vacuum of 2.0 psig.(2) i
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The containment purge supply and exhaust valves are required to be locked closed during plant operations since these valves have not been demonstrated capable of closing from the full open position during a design basis loss-of-coolant accident. Maintaining these valves locked closed during plant operation ensures that excessive quantities of radioactive materials will not be released via the containment purge system in the event of a design basis loss-of-coolant accident. The containment purge supply and exhaust valves will be locked closed by providing locking devices on the control board operators for these valves.
References (1) FSAR - Section 14.3.4 (2) FSAR - Section 5.5.2 15.3.6-3 Unit 1 - Amendment No. 64,86,89,104 i
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The shutdown conditions of the reactor are selected based on the type of activities that are being carried out. When the reactor head is not to be removed, the specified cold shutdown margin of 1% ak/k precludes criticality under any occurrence. During refueling the reactor is subcritical by 5%
ak/k. Positive reactivity changes for the purpose of rod assemb'1y testing l
will not result in criticality because no control bank worth exceeds 3%.
I Positive reactivity changes by boron dilution may be required or small concentration fluctuations may occur during preparation for, recovery from, or during refueling but maintaining the boron concentration greater than 1800 ppm precludes criticality under these circumstances.
1800 ppm is a nominal value that ensures 5% shutdown for typical reload cores. Should continuous dilution occur, the time intervals for this incident are discussed in Section 14.1.5 of the FSAR.
Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal of-coolant accident were as much as 6 psig.II) pressure before a majo The containment is designed to withstand an internal vacuum of 2.0 psig.(2)
The containment purge supply and exhaust valves are required to be locked closed during plant operations since these valves have not been demonstrated capable of closing from the full open position during a design basis loss-of-coolant accident. Maintaining these valves locked closed during plant operation ensures that excessive quantities of radioactive materials will not be released via the containment purge system in the event of a design basis loss-of-coolant accident. The containment purge supply and exhaust valves will be locked closed by providing locking devices on the control board operators for these valves.
References (1) FSAR - Section 14.3.4 (2) FSAR - Section 5.5.2 15.3.6-3 Unit 2 - Amendment No. 69,33,9A,107
. 15.4.4*
CONTAIlOIENT TESTS Applicability t
Applies to containment leakage and stzuetural integrity.
(bioetive To verify that potential leakage from the containment and t.he pre-stressing tendon loads are maintained within acceptable values.
Specification I.
Type A Periodic Integrated 14akage Rate Test A.
h st 1.
The Type A periodic in-service integrated leakage rate test shall be performed at intervals specified in I-C below at an initial pressure P a K e
psig N of des 5 t
pressure (P,)).
2.
Test accuracy shall be verified by supplementary means such as measuring the quantity of air required to return to the starting pressure (P ) or by imposing a known leak t
rate to demonstrate the validity of amasurements.
3.
Closure of the containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves without preliminary exercises or adjustment. Repairs of maloperating or leaking valves shall 15.4.4-1 Unit 1 - Amendment No. M,104 Unit 2 - Amendment No. Ah,107
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be made as necessary. Description of valve closure malfunction or valve leakage that requires corrective action before the test shall be included in the Test Report.
4.
Leak repairs, if required during the integrated leakage test, shall be preceded and followed by. local leakage rate measurements. A description of the repairs and the leakage rates measured prior to and after the repairs shall be included in the Test Report.
5.
The test duration shall not be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the criteria listed in "a" below are met.
a.
For the Absolute Method, Total Time technique, the test duration may be shortened to less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided the following Bechtel Corporation Topical Report (BN-TOP-1) acceptance criteria for short duration testing are met:
(1)
For the containment atmosphere stabilization:
Once the containment is at test pressure the containment atmosphere shall be allowed to be stabilized for about four hours. The atmosphere is considered stabilized when:
1.
The rate of change of average temperature is less than 1.0 F/ hour / hour averaged over the last two hours.
or
- 11. The rate of change of temperature changes Irss than 0.5"F/ hour / hour averaged over the last two hours.
(2)
For the data recording and analysis, using the absolute method, Total Time technique:
1.
The Trend Report based on Total Time calculations shall indicate that the magnitude of the calculated leak rate is trending to stabilize at a value less than the maximum allowable leak rate (L,).
15.4.4-2 Unit 1 - Amendment No./(1 104 Unit 2 - Amendment No. ((, 107
(Note:
The magnitude of the calculated leak rate may be increasing slightly as it tends to stabilize.
In this case, the average rate shall be determined from the accumulated data over the last five hours or last twenty data points, whichever provides the most points. Using this average rate, the calculated leak rate can then be linearly extrapolated to the 24th hour data point.
If this extrapolated value of the calculated leak rate exceeds 75% of the maximum allowable leak rate (L,) then the leak rate test is continued).
and ii.
The end of test upper 95% confidence limit for the calculated leak rate based on Total Time calculations shall be less than the maximum allowable leak rate.
and iii. The mean of the measured leak rates based on Total Time calculations over the last five hours of test or last twenty data points, whichever provides the most data, shall be less than the maximum allowable leak rate, and iv.
Data shall be recorded at approximately equal intervals and in no case at intervals greater than one hour.
and v.
At least twenty data points shall be provided for proper statistical analysis, and vi.
In no case shall the minimum test duration be less than six hours.
B.
Acceptance Criteria 1.
The governing criteria for acceptance of peak pressure tests is that the maximum allowable leakage (L ) shall not exceed 0.40 weight percent per day of containment, atmosphere at 60 psig (P,)
which is the design pressure.
15.4-4 2a J
Unit 1-Amendment No.104 Unit 2-Amendment No.107
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The allowable in-service leakage rate (L ) at the reduced test pressure (P ) shall not exceed L,eq(ukT t,L,(P,/P )1/2. tm/l L /L
), except if L isgreaterfhan0.7,L shall be am Where:
L isthemaxibmallowableleakagerate$tpressurePa forthepfeoperationaltests; the subscript "m" refers to values of the leakage measured during initial preoperational tests; and the subscripts "a" and "t" refer to tests at accident pressure and reduced test pressure, respectively.
The measured leakage rate (Lexceed 0.75 L, as determine Eu)nder B-1 above.
3.
for in-service tests shall not t
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15.4.4-2b Unit 1-Amendnint No. 104 Unit 2-Amendment No. 107
b.
Airlock and equipment door seals, including operating mechanism t
and penetrations with resilient seals which are part of the containment boundary in the airlock structure.
c.
Fuel transfer tube flange seal.
d.
The containment purge cupply and exhaust valves.
e.
Other containment components which require leak repair in order to meet the acceptanc' 41terion for any integrated leakage rate test.
D.
Acceptance criterion 1.
The total leakage from items II.A.5 and III.A.3 shall not exceed 0.6 L,.
If at any time it is determined that 0.6 L, is exceeded, repairs l
a.
l shall be initiated immediately. After repair, a retest to con-firm conformance to the acceptance criterion of :!tI.B. is required.
b.
If repairs are not completed and conformance to the acceptance criterion of 11.3. is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be taken to cold shutdown conditions until repairs are effected and the local leakage meets this acceptance criterion.
2.
The leakage from the airlock doors seal test, resulting from the 3 day testing requirement in II.c.1.d, shall be considered acceptable if the leakage sua from the worst door in each airlock, extrapolated to P,,
and added to the total of items II.A.5 and III.A.3, is less than 0.6 L,.
If the total identified in 11.3.2, above, exceeds 0.6 L,, then the a.
airlock cor:taining the worst door shall be full pressure tested to dotarains the actual inakage perforumee.
3.
The leakage rate for the con +m % t purge supply and exhaust valves shall be compared te the previously unsoured leakage rate to detset excess'ive valve degradation.
C.
' Inst Frequen.cy 1.
Individual penetratices shall be tested during each shutdown for najor fuel reloading except as specified in a and b below. In no case shall the interval be greater thwi two years.
The containment equipsont batch flange seals and the fuel transfer a.
tube flange seals shall be tested at each shutdown for major fuel reloading or after each time used, if that be sooner.
Unit 1 - Amendment No. Af, f6, 104 15.4.4-5 Unit 2 - Amenchent No. 66,1g,107
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b.
The air locks shall be tested at 6-month inter 7als at test pressure not less than P,.
Personnel airlocks shall be tested at a pressure of no less c.
than P, following periods when containment integrity is defeated through the use of the airlock.
d.
Personnel airlocks opened during periods when containment integrity is established shall be tested within 3 days after heing opened. personnel airlocks opened more frequently than once every 3 days shall be tested at least once every 3 days during the period of frequent openings.
The containment purge supply and exhaust valves shall be m.
tested at 6-month intervals.
III.Thpe"C" Tests A Type "C" test measures the leakage across an individual valve or across a group of valves used to isolate an individual penetration through the primary reacter containment as defined in III.A.3.
A.
Test Type "C" tests shall be performed at intervals specified in III.D 1.
below and at a pressure of not less than P,.
~
Acceptable methods of testing are by local pressurization and the 2.
methods described in II.A.4 above. The pressure shall be applied in the same direction as that when the valve would be required to perform its safety function, miess it can be determined that the results from the tests for a pressure applied in a different direc-Each tion will provide equivalent or more conservative results.
valws to be tested shall be closed by normal operation and without any pre 14Mna7 exercising or adjustaants.
f 15.4.4-6 Unit 1 - Amendment No. 61, 16, 104 Unit 2 - Amendment No. 66, /$,107
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3.
Local leakage shall be measured for containment isolation valves that:
a.
Provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal operation.
b.
Are required to close automatically upon receipt of a containment isolation signal.
c.
Are required to cperate intermittently under post-accident conditions.
s B.
Acceptance Criterion The total leakage from items II.A.5 and III.A.3 shall not exceed 0.6 L,.
C.
Corrective Action 1.
If at any time it is determined that 0.6 L, is exceeded, repairs shall be initiated immediately. After repair, a retest to confirm conformance to the acceptance criterion of III.B is required.
2.
If repairs are not completed and conformance to the accept-ance criterien of III.B is not demonstrated within 48 heurs,.
the reactor shall be taken to cold shutdown conditions until repairs are effected and the local leakage meets this acceptance criterion.
D.
Test Frequency 1.
The above tests of the isolation valves shall be conducted during each shutdown for major fuel reloading but in no case at intervals greater than two years.
i 15.4.4-6a Unit 1 - Amendment No. 41,104 Unit 2 - Amendment No. df, 107 l
In addition to the preceding requirements, tesperature readings will E.
be obtained at the locations where inward deformations were measured.
Temperature measurements will also be obtained on the outside of the containment building wall.
g Basis While the reactor The ccatainment is designed for an accident pressure of 60 psig.
operating, the internal environment of the containment will be air at approximately is With these initial conditions, the ctuospheric pressure and a temperature of about 105 F.
temperature of the steam-air mixture at th's peak accident pressure of 60 psig is 286 F.
Prict to initial operation, the containment was strength tested at 69 psig and then 1:ak-tested. The design objective of this pre-operational leakage rate test was estab-lithed as 0.4% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. This leakage rate is consistent with the construction of the containment, which is equipped with independent leak-testable penttrations and contains channels over all containment liner welds, which were indepen-dently leak-tested during construction.
safety analyses have been performed on the basis of a leakage rate of 0.40% by weight With this leakage rate and with minimum contairment engineered per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig.
caf;ty systems for iodine removal in operation, i.e. one spray pamp wita sodium bydr:xide addition, the public exposure would be well below 10 CFR 100 v& lues in the cvent of the design basis accident.
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15.4.4.-12 Unit 1 - Amendment No. g(, 104 Unit 2 - Amendment No. gg, 107
The safety analyses indicate that the containment leakage rates could be slightly in excess of 0.75% per day before a two-hour thyroid dose of 300R could be received at the side boundary.
The perfonnance of a periodic integrated leakage rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic test is to be performed without preliminary leak detection surveys or leak repairs, and containment isolation valves are to be closed in the normal manner. The test pressure of 30 psig or greater for the periodic integrated leakage rate test is sufficiently high to provide an accurate measurement of the leakage rate and it duplicates the pre-operational leakage rate test at 30 psig. The specification provides relationships for relating in a conservative manner, the measured leakage of air at 30 psig or greater to the potential leakage of steam-air mixture at 60 psig and 286*F. The specification also allows for possible deterioration of the leakage rate between tests, by requiring the as measured leak rate to be less than 75% of the allowable leakage rate. The basis for these deterioration allowances are arbitrary judgments, which are believed to be conservative and which will be confirmed or denied by periodic testing.
If indicated to be necessary, the deterioration allowances will be altered based on experience.
The duration of the integrated leak rate test will be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the reduced time duration acceptance criteria are met.
In 1972, the AEC approved a Bechtel Corporation Topical Report, BN-TOP-1, entitlea " Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power." This report provides criteria for short duration testing for the Absolute Method using the Total Time technique. The Bechtel short duration testing criterth contains requirements fer stabilization, leakage rate trending, confidence level, sufficient data for statistical convergence, and allowed leakage rate.
1 The frequency of the periodic integrated leakage rate test is keyed to the refueling schedule for the reactor and shutdown for inservice inspection because these tests can only be performed during refueling shutdowns. The initial core loading was designed for approximately 24 months of power l
operation, thus the first refueling occurred approximately 30 months after initial 15.4.4-13 Unit 1 - Amendment No. 64, 104 Unit 2 - Amendment No. 69, 107
Fesilient seals for these valves.
l References (1) FSAR Section 5.1.2.3 (2) FSM Section 5.1.2 (3) FSAR Section 14.3.5 i
(4) FSAR Section 14.3.4 (5) FSAR Section 6.2.3 (6) FSAR pages 5.1-86 and 5.1-87 l
15.4.4-16 Unit 1 - Amendment No.A9,104 Unit 2 - Amendment No.hti, 107
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