ML20094K004
| ML20094K004 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/11/1992 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-90-06, GL-90-6, ULNRC-2587, NUDOCS 9203170312 | |
| Download: ML20094K004 (8) | |
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, L/dlR)N ILLIUCTHK' f3 Mcrch 11, 1992 U.S.
Nuclear Regulatory Commission Attn:
Document Contrcl Desk Mail Station F1-137 Washington, D.C.
20555
'tlemen:
ULNRC-2 587 DOCKET NUMBER 50-483 CALLAWAY PLANT REVISION TO TECllNICAL SPECIPICATION 2
3/4.4.4. AND_3_.4.9x3
Reference:
ULNRC-2527, dated December 4, 1991 The referenced letter transmitted Union Electric Company's amendment application to address the recommendations of Generic Letter 90-06.
Per our discussions with staff we are submitting additional information to support the significant hazards evaluations previously transmitted.
The attached Significant Hazards Evaluation should be used to replace the previously submitted evaluation.
If there are any questions concerning this matter, please contact me.
Very truly yours,
<Y nff y
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Donald F.
Schnell JMC/plh Attachment 0
9203170312 920311 PDR ADOCK 05000483
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,C.
4 STATE OF MISSOURI i
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Donald F.
Schnell, of lawful age, being first duly sworn upon oath says that he is Senior Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the f acts therein stated are true and correct to the best of his knowledge, information and belief.
By
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ITonald F. Sch'nell Senior Vice President Nuclear SUBSC IBED and sworn to before me this //b-day of
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T. A. Baxter,-Esq.
Shaw, Pittman, Potts & Trowbridge 2300_N. Street, N.W.
Washington, D.C.
20037 Dr.
J. O. Cermak CFA, Inc.
18225-A Flower Hill Way Gaithersburg, MD 20879-5334 R. C.
Knup Chief, Reactor Project Branch 1 U.S. Nuclear Regulatory Commission Region-III 799 Roosevelt Road
-Glen Ellyn, Illinois 60137 Bruce Bartlett Callaway Resident Offit U.S. Nuclear Regulatory Ccmmission RR#1 Steedman, Missouri 65077 h
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R-. Wharton (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E21
-11555 Rockville Pike Rockville, MD 20852 3
Manager,-Electric Department Missouri'Public Service Commission P.O. Box 360 Jefferson City, MO 65102-Ron Kucera.
Department of Natural Resources P.O. Box 176-Jefferson. City,.MO 65102 l
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/QA Record (CA-758)
Nuclear Date E210.01 DFS/ Chrono D.
F.
Schnell J.
E.
Birk J. V.
Laux M.
A.
Stiller G.
L.
Randolph R.
J.
Irwin P.
Barrett C.
D. Naslund A.
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Passwater D.
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Shafer W.
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Kahl S. Wideman (WCNOC)
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Attachment Page 1 of 4 UbNRC-2587 SIGNIFICANT HAZARDS EVALUATION Th.is amendment application requests a change to Technical Specifications (T/S) 3/4.4.4 on Relief Valves and its associated Bases and T/S 3.4.9.3 on Overpressure Protection System.
These specifications are being revised to incorporate certain NRC staff 9
positions resulting from the resolution of Generic Issue 70 and 94 as presented in Generic Letter (GL) 90-06.
The proposed changes to T/S 3/4.4.4 are as follows:
1.
ACTION statement a is being revised to include the requirement to malatain power to closed block valve (s) because removal of power would render the block valve (s) inoperable and the requirements of ACTION d. would apply, Power is maintained to the block valve (s) so that it is operable and may be subsequently opened to allow the PORV to be used to control reactor coolant system pressure.
Closure of the block valve (s) establishes reactor coolant pressure boundary integrity for a PORV that has excessive seat leakage.
Reactor coolant pressure boundary integrity takes priority over the capability _of the PORV to mitigate an overpressure event.
2.
ACTION statements a.,
b.,
and c. are being changed to terminate the forced shutdown requirements with the plant being in HOT SHUTDOWN rather than COLD SHUTDOWN because the APPLICABILITY requirements of the LCO do not extend past the HOT STANDBY mode.
3.
ACTION statement d.
is being changed to establish remedial Q
measures-that are consistent with the function of the block valves.
The prime importance for the capability to close the block valve is to isolate a stuck-open PORV.
Therefore, if the block valve (s) cannot be restored-tc operable atatus within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the remedial action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck-open PORV at a time that the block valve is inoperable.
The modified ACTION statement.does not specify closure of the block valves because such action would not likely be possible when the block valve is inoperable.
.t does not specify either the closure of the
- Likewise, i
PORV, because_it would not likely be open, or the removal of power from the PORV.
When the block valve is inoperable, placing the PORV in manual control is sufficient to preclude the potential for having a stuck-open PORV that could not be isolated because of an inoperaole block valve.
The proposed changes to Technical Specification 3/4.4.4 and its Bases-do not involve a significant hazards consideration because operation of Callaway Plant with these changes would not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated.
No credit l
1
Attachment Page_2 of 4 ULNRC. 2 58 7 is taken for operation of the PORVs in the PSAR Chapter 15 accident analyses if their operation mitigates the result of the accident.
Turbine trips are evaluated in FSAR Section 15.2.3 with and without the pressurizer PORVs.
The loss of of f site AC power and loss of normal feedwater analyses (FSAR Sections 15.2.6 and 15.2.7) assume the PORVs are operable only because their operation neximizes the transient
-pressurizer water volume caused by condensation of steam that would'have been relieved through the safety valves.
The proposed changes to Technical Specification 3/4.4.4 Action a.
requires that with the block valve (s) closed, power be maintained to the valve (s) _sc they can be readily opened from the control room.
This change would decrease the amount of time needed to initiate feed and bleed capabilities in the event an alternative measure to remove decay heat from the reactor core is necessary.
The proposed change to T/S 3/4.4.4 Action d.
is a clarification for a potential situation where an automatic signal to the PORVs is inoperable but the PORV is mechanically functional.
Since the PORV is still mechanically functional, it would enhance safe operation to not close and remove power from the block valve, and allow the PORV to remain in a condition where it could easily be manually opened from the control room if required.
This clarification is consistent wich the operability requirements for the PORVs in Modes 1, 2 and 3.
Therefore, the proposed changes to Technical Specification 1/4.4.4, and its associated Bases are intended to increase the reliability and availability of the PORVs, and do not involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
Create the possibility of a new or difterent kind of accident from any previously evaluated.
There is no new type of accident or malfunction being created and the-method and_ manner of_ plant operation remains unchanged.
No change in testing methodology is_being proposed, and the equipment is not being operated in a new or different manner. -Changes incorporate the staff positions delineated in GL 90-06.
3.
Involve a significant reduction in margin of safety.
There are 1. ' plant _ design changes involved and no changes are bein; ns le to ' the safety limits or safety system settings that would adversely impact plant safety. -The proposed changes to Technical Specification 3/4.4.4 increase the avai-lability and reliability of the power-operated relief valves (PORVs) and block valves to perform their intended i
function.
The changes do not reduce any technical specification margin of safety.
The proposed changes to T/S 3.4.9.3 are as follows below:
1.
The LCO statement is being modified to require that at least two overpressure protection devices must be operable.
That is, two PORVs or two Residual Heat Removal ( RHR.) Suction relief valves or one PORV and one RHR suction relief valve
Attachment Page 3 of 4 ULNRC-258 7 l
must be operable when cold overpressure protecti required.
At Callaway, on is RHR suction relief valves,the operability of two PORVs, and one PORV, or a reactor coolant system ventof one RHR suction relief v of two least 2 square inches ensures that opening of at as required by 10CFR50, Appendix G. the RCS will be protected 2.
ACTION statement a.
applicable in Modes 3 or 4.is revised to clarity that it At Callaway is only protection system is required to be opera,ble in M dthe overpressure 368 F and in Modes 4, 5, and 6.
o e 3 below statement makes it This revision to the ACTION one pressure relief device is operable.only applicable in Modes 3 or 4 wh with GL 90-06 Attachment B-1.
This is consistent 3.
ACTION Etatement b.
for one of the two required PORVs or RHRis adde time (ACT) relief valves to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in Modes 5 or 6 suction or 2 RHR Suction relief valves or 1 That is, 2 PORVs protection is required. relief valve must be operable when cold overp PORV and 1 RHR suction ressure conditions under which a low-temperature overThe NRC has considered transient is most likely to occur.
pressure the most vulnerable period of time waoverpressure protect shutdown modes, with the reactor coolant s found to be Mode 5 200 F, especially when water solid, temperature less than or equal to evaluation of operating reactor experiences perfbased on th_ detailed support of CI 94.
The staff concluded that ormed in low-temperature overpressure protection sy t t
safety-related function and inoperable overpre s em performs a protection equipment should be restored to an o ssure status in a shorter period of time.
perable is considered by the NRC to be too long und The current 7-day AOT conditions.
The NRC has concluded that ar certain potential for an overpressure transientreduced to 24 hou or 6 when the is highest.
The proposed changes to Technical Specifi involve a significant hazards conside cation 3.4.9 Callaway Plant with these changes wouldration because op.3 do not eration of not:
1 Involve a significant consequences of an accident previously evaluatedincrease in th no change being proposed in the control design d There is the occurrence of an overpressure transi e
to limit changes to Technical Specification 3.4 ent.
The proposed limit the amount of time the plant
.9.3 only serve to potentially danaging overpressure transient with liis vulnerable to a overpressure protection available.
mited changes to Technical Specification 3.4Therefore, the proposed flexibility and availability of the low t 9.3 increases overpressure protection system with a result
- emperature ant increase in
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u Attachment.
Page 4 of 4
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ULNRC-2 587
- the level of plant safety and do not involve a significant
~
increase in the probability or consequences of an accident previcusly evaluated,
'2.
Create-the_ possibility of a new or different kind of accident from any previously evaluated.
There is no new type of-accident or malfunction being created and the method and manner of plant operation remains unchanged.
There are no changesfbeing proposed to the level of surveillance required to demonstrate compliance with the LCO.
The installed overpressure mitigation devices will continue to be operated and tested in a. manner consistent with their
- design-and-installation.
The proposed changes are intended to enhance the level of overpressure protection during periods of vulnerability.
Changes jncorporate the staff positions-delineatedEin GL 90-06.
13.
Involve a significant reduction in.a margin of safety.
There are no plant design changes involved and no changes are being made to'the' safety limits or' safety system settings that would adversely impact plant safety.
The proposed-changes to Technical Specification 3.4.9.3 increases the flexibility and availability of the overpressure protection system to mitigate a low-temperature cverpressurization event.
The changes do not reduce any i
technical specifications margin of safety.
l Based on the above discussions, it has'been determined that the Jrequested Technical Specification revisions do not involve a
~significant increase in the probability or consequences of an L
. accident;or other adverse condition over previous evaluations; or create-the possibility of a new or:different kind of accident or condition over previous evaluations; or involve a-significant reduction in a margin of-safety.
-Therefore, the requested license amendment does not involve a significant hazards iconsideration.
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