ML20003H286
| ML20003H286 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 04/20/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20003H200 | List: |
| References | |
| NUDOCS 8105050487 | |
| Download: ML20003H286 (7) | |
Text
.
ATTACHMENT 3 ATTACHMENT TO ORDER FOR MODIFICATION OF LICENSE DATED APR 2 0 1091 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285
, Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages contain vertical lines indicating the area of change.
Pages 2-2b 2-2c 2-2d 3-25 3-26 3-29 j
l i
l6[b b D
2.0 LIMITING CONDITIONS FOR O'?ERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Ooerable Comnonents (Continued)
(a) A pressurizer steam space of 60% by volume or greater exists, or (b) The steam generator secondary side temperature is less than 50 F above that of the reactor coolant system cold leg.
(12) Reactor Coolant System Pressure Isolation Valves (a) The integrity of all pressure isolation valves listed in Table 2-9 shall be demonstrated, except as specified in (b). Valve leakage shall not exceed the amounts indicated.
(b) In the event that the integrity of any pressure isolation valve specified in Table 2-9 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition.*
(c)
If Specifications (a) and (b) above cannot be met, an 1
orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Basis The plant is designed to operate with both reactor coolant loops and t
associated reactor coolant pumps in operation and maintain DNBR above t
1.30 during all normal operations and anticipated transients.
In the hot shutdown mode, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be operable.
In the cold shutdown mode, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing l
decay heat, but single failure considerations require that at least I
two loops be operable. Thus, if the reactor coolant loops are not operable, this specification requires two shutdown cooling pumps to be operable.
The requirement that at least one shutdown cooling loop be in operation during refueling ensures that:
(1) sufficient cooling capacity is avail-able to remove decay heat and maintain the water in the reactor pressure vessel below 210 F as required during the refueling mode, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.
~
Manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supplies deenergized.
[
Amendment No. $6, Order dated 4/20/812-2b
2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)
The requirement to have two shutdown cooling pumps operable when there is less than 15 feet of water above the core ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 15 feet of water above the core, a large heat sink is available for core cooling; thus, in the event of a failure of the operating shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the core.
When reactor coolant boron concentration is being changed, the process must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the reactor coolant is assured if one low pressure safety injection pump or one reactor coolant pump is in operation. The low pressure safety injection pump will circulate the reactor coolant system volume in less than 35 minutes when operated at rated capacity. The pressurizer volume is relatively inactive; therefore, it will tend to have a boron concentration higher than the rest of the reactor coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer spraystomaintainanominalspreadbetweentheboronconcentrationintgg) pressurizer and the reactor coolant system during the addition of boron.
Both steam generators are required to be filled above the low steam gener-ator water level trip set pcint whenever the temperature of the reactor coolant is greater than the design temperature of the shutdown cooling j
system to assure a redundant heat removal system for the reactor.
I The design cyclic transients for the reactor system are given in FSAR Section 4.2.2.
In addition, the steam generators are designed for addi-tional conditions listed in FSAR Section 4.3.4.
Flooded and pressurized conditions on the stean side assure mintzum tube sheet temperature differential during leak testing. Themgnimumtemperatureforpressuriz-ing the steam generator steam side is 70 F.
Formation of a 60% steam space ensures that the resulting pressure increase would not result in an overpressurization, should a reactor coolant pump be started when the steam generator secondary side temperature is greater than that of the~ RCS cold leg.
I For the case in which no pressurizer steam space exists, limitation of the steam generator secondary side /RCS cold leg aT to 50 F ensures that a single low set point PORV would prevent an overpressurization due to actuation of a reactor coolant pump.
I f
The exception to Specification 2.1.l(4) requiring all containment penetra-(
tions providing direct access from the containment to the outside atmos-phere be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requires that the equipment hatch be closed and held in place by a minimum of four bolts.
References (1) FSAR Section 4.3.7 A:endment No. 77, Order dated 2-2c BQN)fR
I TABLE 2-9 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum (a)(b)
(
System Valve No.
Allowable Leakane l
High-Pressure Safety Injection Loop 1A, Cold leg SI-216 3,5 gpm SI-201
- _5 gpm l
Loop 1B, cold leg SI-220 5,5 gpm SI-204 5,5 gym Loop 2A, cold leg SI-208 3,5 spm SI-195 5,5 gpm Loop 2B, cold leg SI-212
- ,5 gpm SI-198
- _5 gym Low-Pressure Safety Injection Loop 1A, cold leg SI-200 5,5 spm Loop 1B, cold leg SI-203 I 5 spm Loop 2A, cold leg SI-194
- _5 gpm Loop 23, cold leg SI-197
< 5 gym Footnotes:
(a) 1.
Leakage rates less than or equal to 1.0 gpm are considered acceptable.
2.
Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
3.
Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm i
are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the i
margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
4.
Leal; age rates greatar than 5.0 gpm are considered unacceptable.
(b) Minimum test differential pressure shall not be less than 150 psid.
Order /ated 4/20/81 2-2d
,,,__,---..v.,
.-_.-__..,.my-.
--.._,_---.,--..--...-,r--,,
,,._,,-,,n.y,, _. -,,. _ ~ _
,,,..,., -. ~ -., -. - -.. -,. _. _. - -.
3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant System. Steam Generator Tubes. and Other Components Subiect to ASME XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance (Continued)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.
(ii)
The steam generator shall be determined OPERABLE
.after completing the corresponding actions-(plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 3-8.
e.
Reporting Requirements (i)
Following each in-service inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 30 days.
(ii)
The complete results of the steam generator tube inservice inspection shall be reported to the Commission within six (6) months following com-pletion of the inspection. This report shall include:
j 1.
Number and extent of tubes inspected.
l 2.
Location and percent of wall thickness pene-l tration for each imperfection.
3.
Identification of tubes plugged.
(iii) Results of steam generator tube inspections which fall into Category C-3 and require prompt notifica-tion of the Commission shall be reported pursuant to Section 5.9.2 of the Technical Specifications prior to resumption of plant operation. The written follow-up of this report shall provide a description of inves-tigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
(3) Surveillance of Reactor Coolant System Pressure Isolation Valves a.
Periodic leakage testing
- on each valve ' ;ted in Table 2-9 shc11 be accomplished prior to entering che power operation
- To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that tha cathod is capable vf demoustrating valve compliancs with tha leakage criteria.
Amendment No. 46, Order dated 3-25
/MSM8Jt
3.0 SURVEILLANCE REQUIRDENTS 3.3 Reactor Coolant System. Steam Generator Tubes. and Other Components i
Subiect to ASMI XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance (Continued) mode every time the plant is placed in the cold shutdown condition for refueling, each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, and prior to return-ing the valve to service after maintenance, repair or re-placement work is performed.
b.
Whenever the integrity of a pressure isolation valve listed in Table 2-9 cannot be demonstrated the integrity of the remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other valve located in the high pressure line shall be recorded daily.
i j
l l
l l~
(
l Amendment ilo. 16, Order dated 3-26 l
4/20/21
l 3.0 SURVEILLANCE REQUIREMENTS 3.3 Reactor Coolant System. Steam Generator Tubes and Other Components Subiect to ASME XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance (continued)
Basis Undetected prolonged leakage of borated reactor coolant onto carbon steel sets up an unusual corrosion mechanism. Detection of this leakage at an early stage can best be accommodated directly after an outage and before startup. The inspection program specified in Specification 3.3(1) places major emphasis on the areas of highest stress concentration as determined by general design evaluation and experience with similar systems. The inspections will rely on non-destructive analysis methods utilizing up-to-date analyzing equipment and trained personnel. Volumetric inspection of the reactor presrure vessel is to be performed completely from the outside diameter. The testing techniques and acceptance criteria of Section XI of the ASME B&PV Code will be utilized, except where specific relief is granted by the Commission.
The surveillance requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for in-service inspection of the steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1, dated July 1975. In-service inspection of steam generator tubing is essential in order to maintain surveillcace of the conditions of the tubes in the event that there is evidence of mechanical damage or progres-sive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion.
In-service inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even is a defect should develop in service, it will be found during scheduled in-service steam generator tube examina-tions. Plugging will be required for all tubes with imperfections exceed-ing the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
f Whenever the results of any steam generator tubing in-service inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Section 5.9.2 of the Technical Specifications prior to the resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for 1
analysis, laboratory examinations, tests, additional addy-current inspec-tion, and revision of the Technical Specificacions, if necessary.
References (1) FSAR, Section 4.5.3
/
Amendment No. 46, Order dated 4/20/21 3-29
-*+=n e
-ar w-n, e-w--mwavs--+-
w- - -
mw-e---.
--,--rv.-e--w
- w. - - - -
-+-m,-w,-
n-+v-
-a---
+s+
r<-----+-----v-,m.-w--ma*r r me -i wi--+=q*w-u-----vs ye t-e'wr+-
a'*wv-*v+
v-er