ML20071A715
| ML20071A715 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/15/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20071A717 | List: |
| References | |
| NUDOCS 8302240337 | |
| Download: ML20071A715 (6) | |
Text
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J' TS.3.1-3 a.
7 Basis When the boron concentration of the reactor coolant system is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.. The residual hear removal pump will circulate the equivalent of the primary system volume in app 'ximately one-half hour.
" Steam Generator Tube Surveillance", Technical Specification 4.12, identifies steam generator tube imperfections having a depth >50% of the 0.050-inch tube wall thickness as being unacceptable for power operation. The results of steam generator burst and tube collapse tests submitted to the staff have demonstrated that tubes having a wall th!ckness greater than 0.025-inch have adequate margins of safety against failure due to loads imposed by normal plant operation and design basis accidents.2 Part A of the specification requires that both reactor coolant pumps be operat-ing when the reactor is critical to provide core cooling in the event that a loss of flow occurs.
In the event of the worst credible coolant flow loss (loss of both pumps from 100% power) the minimum calculated DNBR remains well above 1.30.
Therefore, cladding damage and release of fission products to the reactor coolant will not occur. Critical operation, except for low power physics tests, with less than two pumps _is not planned. Above 10% power, an automatic reactor trip will occur if flow from either pump-is lost. Below 10% power, a shutdown under administrative control will be made if flow from either pump is lost.
The pressurizer is needed to maintain acceptable system ? essure during normal plant operation, including surges that may result following anticipated transients. Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point. Below 350*F and 450 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby contro1 ' system temperature and pressure.
If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less
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than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization of the reactor coolant system for reactor coolant temperatures less than 350*F.
The combined capacity of both safety valves is greater than the maximum surge rate resulting from complete loss of load.1 7_ _ __
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Prairie Island Unit 1 AmendmentNo.47,gg,61 Prairie Island Unit 2 Amendment No. 41, 41, 5 5 I
B302240337 830215 PDR ADOCK 05000282 P
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TABLE TS.4.1-1 (Page 3 of 5)
Channel Functional
Response
3 Description Check Calibrate Test Test Remarks e
16.
Refueling Water W
R M(1)
NA
- 1) Functional test can be performed Storage Tank Level by bleeding transmitter 17.
Volume Control Tank S
R NA NA 18a. Containment Pressure S
R M(1)
NA Wide Range Containment Pressure
{
SI Signal
- 1) Isolation Valve Signal U
18b. Containment Pressure S
R H
NA_
Narrow Range Containment Pressure Steam Line Isolation o
18c. Containment Pressure S
R H
NA l2 Containment Spray ll 18d. Annulus Pressure (Vacuum Breaker)
NA R
R NA e
- 19. Deleted u
E 2D. Boric Acid Hake-up Flow NA R
NA NA Channel w
21.
Containment Sump Level NA R
R NA Includes Sumps A, B, and C 5
22.
Accumulator Level S
R R
NA k
s and Pressure g-S 23.
Steam Generator Pressure S
R H
HA m
g 24.
Turbine First Stage Pressure S.
R M
NA
.-t 2"
- 25. Emergency Plan Radi*ation
- M R
H NA Includes those named in the emergency sa Instruments procedure (referenced in Spec. 6.5 A.6)
L.
g ba r.
cn
,26.
Protection Systems NA NA M
NA Includes auto load sequencers Logic Channel Testing dy w
O Nt
- r
ll TAnts TS.4.1-1 7
Page'4'of 5)
Functional
Response
Channel Description Check Calibrate-Test Test
. Remarks I
- 27. Turbine Overspread
' NA R'.
H NA 4,
l'/
Protection Trip Channel
>7-w i
4 28.
Deleted 4,*
f 29.
Deleted
- r 9
2, 30.
Deleted a
y 31.
Seismic Honitors R
R NA NA Includes those reported in Item 4 of Table TS.6.7-1 5
1*
,3 2.
Coolant; Flow - RTD S
R H
HA Bypass Flowmeter
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- 33. CRDH Cooling Shroud S
NA R
NA FSAR page 3.2-56 Exhaust Air Temperature
~
34.
Reactor Gap Exhaust S
NA R
HA FSAR page 5.4-2 Air Temperature
- 35. Post-Accident Honitoring H
R NAl NA Includes all t' hose in FSAR Table e
S, 7.7-2 and Table TS.3.15-1 not in-cluded elsewhere in 'this Table t
n s
I
- 36. Steam Exclusion W
R' H
NA y
Actuation System i
~ See FSAR Appendix I,Section I.14.6 g
y-
- 37. Overpressure NA R
R NA-Instrument Channels for PORV g
g Hitigation System Control Including Overpreisure g
a Hitigation System M-i o
a 38.
Degraded. Voltage NA R
H NA 4KV Safeguard Busses
,ta
~,
y
- 39. L'oss of Voltage NA R
H NA wm 4KV Safeguard Busses
^
M
- G M
y g
v e
i.
i
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l TABLE TS.4.12-1 STEAM GENERATOR TUBE INSPECTION
~
IST SAMULE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE THEPECTTON Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.C.
C-2 Plug defective tubes and inspect additional C-1 None N/A N/A 2S tubes in this S.C.
C-2 Plug defective tubes C-1 None and inspect additional C-2 Plug defective tubes 4S tubes in this S.G.
C-3 Perform action for C-3 result of first sample C-3 Perform action for N/A N/A C-3 result of first sample C-3 Inspect all tubes in All. other -
None N/A N/A this S.G., plug de-S.G.s are fective tubes and C-1 inspect 2S tubes in Some S.G.s Perform action for N/A N/A each other S. G.
C-2 but no C-2 result of second H
5 additional sample N.
Prompt notification S.G. are to NRC.
C-3
.g Additional Inspect all tubes in N/1 N/A
- 'p S.G. is C-3 each S.G. and plug defective tubes.
j' Prompt notification to NRC.
S=3%; When two steam generators are inspected during that' outage.
S=6%; Nhen one steam generator is inspected during that outage.
^
Unit 1 - Amendment No. 31, G 1 i
Unit 2 - Amendment No. 25. 5 5 7
TS.6.7-3a 7
Report of Safety and Relief Valve Failures and Challenges. An annual report of pressurizer safety and relief valve failures and challenges shall be submitted prior to March 1 of each year.
B.
Reportable occurrences Reportable occurrences, including corrective actions and measures to 4
prevent recurrence,. shall be reported to the NRC. Supplcmantal reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be.ompleted and reference shall be made to the original report date.
Unless explicitly stated, the requirements of this section do not apply to the fire protection systems and measures contained in Sections 3.14/4.16, the radiological effluent limitations and measures in Sections 3.9/4.17, or the radiological environmental monitoring program in Section
'4.10.
Fire protection reporting requirements have been separately j
specified in those sections.
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Unit 1 - Amendment No. 9, 59, 61 i
unit 2 - Amendment No. 4, 53, 55 i.
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i 5-TS.6.7-4 l
1.
Prompt Notification With Written Followup. The types of events listed below shall be reported as expeditiously as possible, but
~ within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Administrator of the appropriate Regional NRC Office or his designate no later than the first work-ing day following the event, with a written followup report within two weeks. The written followup report shall include, as a minimum, a completed copy-of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of.the circumstances surrounding the event.
(a) Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored. parameter reaches 'the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.
Note: Instrement drif t discovered as a result of testing need not be reported unde'r this item but may be reportable under items B.l(e), B.1(f), or B.2(a) below.
(b) Operation of the unit or affected systems when any parameter or operation subject to a limiting condition is less conserv-ative than the least conservative aspect of the limiting condi-l tion for operation established in the technical specifications.
1 Note: If specified action is taken when a system is found to be operating between the most conservative and the least conservative aspects of a limiting condition for operation listed in the technical specificationa, the limiting condi-tion for operation is not considered to have been violated and need not be reported under this item, but it may be reportable under ' item B.2(b) below.
(c) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.
3 Note: Leakage cf valve packing or gaskets within the limits for identified leakagn set forth in technical speciffcations i
need not be reported under this item.
I Unit 1 - Amendment No. 9,///, 61 Unit 2 - Amen' ament No. 4, yd, 55
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