ML20071N300
| ML20071N300 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/24/1982 |
| From: | Jaffe D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20071N297 | List: |
| References | |
| NUDOCS 8210070125 | |
| Download: ML20071N300 (4) | |
Text
_.
eMec%
~%
UNITED STATES **
g NUCLEAR REGULATORY COMMISSION
{
3 E-WASHINGTON, D. C. 20555
\\...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NOS. 75 AND 56 TO -
FACILITY OPERATING LICENSES NOS. DPR-53 AND DPR-69 BALTIM0RE GAS AND ELECTRIC COMPANY i
CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS. 1 & 2 DOCKET NOS. 50-317 AND 50-318 Introduction By application for license amendments dated August 2, 1982, Baltimore Gas and Electric Company (BG&E) requested changes to the Technical Specifications (TS) for Calvert Cliffs Units 1 and 2.
The proposed changes address.the following:
(1) typographical errors are corrected involving inservice inspection (TS 4.5.2.g.2) and valve testing (TS 3.4.3.c); (2) a change to TS 3.4.3.c to allow startup of the reactor with one or more inoperable, but isolated. power operated relief valves (PORVs); (3) changes to several TS to incorporate numerical values of containment leakage rates La and Lt; and (4) a change to TS 6.8.1.a.to incorporate by reference the latest applicable version of Regulatory Guide 1.33, " Quality Assurance Program Require-ments (Operation)".
Discussion and Evaluation 1.
Correction of Typographical Errors TS4.4.10.1.2,"AugmentedInserviceInspectionProgramforMain Steam and Main Feedwater Piping", requires an "...augnidnte'd~ inservice inspection program using the applicable rules, acceptance criteria, and repair procedures of the ASME Boiler and Pressure Vessel Code,Section XI, 1974 Edition and Addenda through Summer 1976, for Class 2 components.'.' A review of the. inservice inspection program require-ments for Calvert Cliffs indicates that the Summer 1976 Addenda is
. not applicable to Calvert Cliffs. The correct reference is to the Summer 1975 Addenda. Accordingly, it is appropriate to correct TS 4.4.10.1.2 by changing "... Addenda through Summer 1976", to
... Addenda through Summer 1975".
The second typographical error corrected herein involves TS.4.5.g.2,
" Emergency Core Cooling Systems". TS 4.5.g.2 requires the verifica-tion of the correct position of the auxiliary high. pressure-safety injection (HPSI) loop isolation valves. One such valve, MOV-646, is ryeated twice in the list of valves to be checked. A review of drawing M-74, "Calvert Cliffs Safety Injection and Containment Spray System" indicates that auxiliary HPSI loop isolation valve MOV-647 DESIG ED ORIGINAI.
f
&#AL
/N' 8210070125 820920 CortifiedBFj-PDR ADOCK 05000317 P
\\
is missing from the list of valves in TS 4.5.g.2..From the above, we conclude that one of the duplicate entries for valve MOV-646 was,
~
- most likely, intended to be for valve MOV-647. Accordingly, it is i
appropriate to change one of the entries for valve MOV-646 to an entry for valve MOV-647 in TS 4.5.g.2.
2.
Startup of the Reactor with an Isolated PORV The pressurizers for Calvert Cliffs 1 and 2 are each equipped with two ASME Code approved safety valves and two power operated relief valves (PORVs).
Each PORV is capable of.being. isolated by its associated i
block valve. The Code Safety Valves, alone, are sufficient to pro-vide over pressure protection from the most severe pressure transient initiated from full power. At the present time, TS 3.4.3, " Reactor Coolant Relief Valves", allows full power operation with failed PORV(s) provided that.the associated. block valves are closed and-electrical power removed for the valve..Startup of.the reactor with isolated PORVs is not permitted per TS 3.0.4*, however, since a.TS
" Action Statement" would have been in.itiated. By application dated August 2, 1982, BG&E has requested that the reactor be allowed to start up, thereby changing operation modes, with inoperable PORVs that have been properly isolated.by closing their associated block valve (s) an.d removing electrical power to the valve (s).
As indicated above, full power operation with failed PORVs that are.
' properly isolated is permited'under TS 3.4.3..The Code Safety Valves i
are sufficient to protect the reactor coolant system under the most severe, full-power, overpre';sure, conditions.** Since there are no safety considerations associated with the startup of the. reactor with isolated PORV(s) more r,estrictive than those already considered for' full power operation it is appropriate to allow startup of the. -
l reactor with isolatoi PGRV(s). Accordingly, the following words' should be-added to TS 3.4.3 as item "c":
"With one or more block-valve (s) closed, and power removed from the block valve (s) to satisfy a or b.above,-the provisions of Specificati.on 3.0.4 are not applicable."
In addition to the above change / the licensee.has requested relief'from
(
~.
TS 4.4.3.2 as it relates to PORY block valve testing...Theilicensee has requested that, in the event that the PORV block valve'has been closed and isolated in response to an inoperable PORV, block valve testing.should be suspended. We concur in the licensee's request in that block valve testing with subsequent failure could prevent the isolation of a failed PORV. Accordingly, we find this change acceptable.
..i.___._.__....... _ _ _..... m.._ _. _.. _ _.
- . TS 3.0.4 does not permit changing operational modes when a TS
" Action Statement" has been initiated.
TS 3.4.3 applies only to operation. modes 1, 2 and 3 (power opera-tion, startup and hot standby). Overpressure protection with the reactor coolant temperature below.2750F (modes 4, 5 and 6)"i'siddressed sep.arately in TS 3.4.9.3 and is not affected by this proposed TS change'.
l i.
-~
-_ U 3.
Numerical Values of La and Lt By letter dated July 22, 1982, BG&E forwarded LER 82-22/IT, Revil-sion 1.. This LER reported an administrative inadequacy whereby a personnel error, during revision of containment local leak rate surveillance test procedures, resulted in the total Type B* and C*
containment leak rate to be compared to La* rather than.6 La. The corrective measures proposed in LER 82-22/IT, Revision 1, included' the introduction of nume.rical values for La and Lt* into TS 3/4 6.1.2,
" Containment Leakage", 3/4.6.1.3, "Coatainment Air Lock", and 4.6.4,
Containment Isolation Valves".. The introduction of these numerical values, and fractions of these values as appropriate, would help assure correct comparison of measured leakage rates with limiting values in the TS.
The addition of' numerical values of La and Lt to the TS, and fractions I
of these values as appropriate, in no way changes the associated requirements. The introduction of these numerical values is'adnfinistra-c tive in nature-and has no effect on the safety of the facility.
i 4.
Updated Reference to " Appendix A" of Regulatory Guide 1.33 Appendix A to Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)" was part of the original licensing basis
- for Calvert Cliffs Units 1 and 2.
The requirement to establish, implement and maintain procedures in accordance with. Appendix A of Regulatory Guide 1.33 is specified in TS 6.8.la.
The commitment to Appendix A'of Regulatory Guide 1.33 was subsequently updated to I
include Appendix A of Regulatory Guide 1.33, Revision 2 February 1978.
By application dated August 2, 1982, BG&E requested a ch'ange
~.
1 to TS 6.8.la to reflect their updated commitment.
Appendix A to Regulatory Guide 1.33, Revision 2, represents the latest approved guidance from the NRC regarding the safety related activities to be controlled by procedures. This document is referenced in the BG&E Operational Quality Assurance Program as described in Section I.B.2.of the'Calvert Cliffs Updated Final Safety Analysis Report (FSAR). Accordingly, it is appropriate to change.TS 6.8.la to in-corporate Appendix A t'o' Regulatory Guide 1.33, Revision'2,.by refer-ence.-
Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of l
environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an l
environmental impact statement or negative declaration and environ-
[
mental impact appraisal need not be prepared in connection with thei l
issuance of these amendments.
- These terms are defined in 10 CFR Part 50, Appendix J, Section II.A.
i,,,_,,,. -..
4
--.-+e Sr
-"-&---4' "r-*---'
4 o
4-Conclusion We have concluded, based on the considerations discussed above, that:
I' (1) because the amendments do not involve a sigr.ificant increase in tt; probability-or consequences of an accident previously evaluater, do not create the possibility of an accident of a type different from any evaluated previously, and do not involve a significant reduction in a margin of safety, the amendments do'not involve a significant hazards consideration, (2) there is reasonable assurance that the health l
and safety of the public will not be endangered by operatior, in the proposed manner, and (3) such activities will' be conducted in compliance
-with the Commission's regulations and the issuance of the amendments _will not be inimical to the common defense and security or to the health and safety of the public.
Date:
September 20, 1982 l'
Principal Contributor:
l t
Dave Jaffe j
t 5
I i
i i
I f
l e
0 0
e v.
-