ML20148C962
| ML20148C962 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/20/1997 |
| From: | Bowling M NORTHEAST NUCLEAR ENERGY CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20148C968 | List: |
| References | |
| B16471, GL-91-08, GL-91-8, TAC-M94623, NUDOCS 9705290314 | |
| Download: ML20148C962 (18) | |
Text
. --_
_m
. _._._-.__._ _- ~
m._____
I g
i
'Y #
Northeast
"P" '"Y 'id- ("""'" '56)' **"d"'d C' "385 i
- Nuclear Energy staistone Nucica, powe, siation i
Northeast Nuclear Energy Company P.O.11ox 128
-j Waterford, CT 06385-0128 (860) 447-1791
]
8 Fax (860) 444-4277 The Northeast Utilities System i
MAY 20 9JT t
Docket No. 50-336 B16471 l
Re: 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk i
Washington, DC 20555 i
i Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Containment Isolation Valves (TAC No. M94623)
Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO), in a letter 3
3 dated January 22,1996, proposed to amend Operating License DPR-65 by changing l
the Technical Specifications of Millstone Unit No. 2. The proposed changes modified Definition 1.8 " Containment Integrity," Technical Specification 3.6.1.1 " Containment Integrity," Technical Specification 3.6.3.1 " Containment Isolation Valves," and the Bases for Technical Specification 3.6.3.1 " Containment isolation Valves." The i
proposed change relocated the list of containment isolation valves, Table 3.6-2, from the Technical Specifications to the Technical Requirements Manual, in accordance with NRC Generic Letter (GL) 91-08', which provides guidance for removal of component L
lists from the Technical Specifications.
A review of the original submittal revealed deviations from the guidance contained in l
GL 91-08. These deviations are not necessary. Therefore, NNECO is proposing to rectify this by revising the original submittal, dated January 22, 1996, with this 4
submittal. This submittal is consistent with the recommendations contained in GL 91-08.
- provides the discussion of proposed changes. Attachment 2 contains the Significant Hazards Consideration. Attachment 3 provides the new marked-up version of the appropriate pages of the current Technical Specifications.
I provides the retyped pages of the Technical Specifications. Attachment 5 provides the i
IN j
J.G. Partlow letter to All Holders of Operating Licenses or Construction Permits i
for Nuclear Power Reactors, " Removal of Component Lists from Technical i l
Specifi gpneric Letter 91-08)," dated May 6,1991.
hlll.ll.lellclhlllllllll 4705290314 970520 PDR ADOCK 0500o336 P
,.4 U.S. Nucle:r Rngulatory Commission B16471/Page 2 original marked-up version of the appropriate pages of the current Technical Specifications.
The proposed changes were evaluated utilizing the criteria of 10CFR50.59, and were determined not to involve an unreviewed safety question.
Additionally, we have concluded that the proposed changes are safe.
The proposed changes do not involve a significant impact on public health and safety (see the Safety Assessment provided in Attachment 1) and do not involve a Significant Hazards Consideration pursuant to the provisions of 10CFR50.92 (see the Significant Hazards Consideration provided in Attachment 2).
Environmental Considerations NNECO has reviewed the proposed license amendment request against the criteria of 10CFR51.22 for environmental considerations. The proposed changes relocate the list of containment isolation valves from the Technical Specifications to the Technical Requirements Manual and reduce the surveillance frequency to verify the position of containment isolation valves located inside containment.
These changes do not increase the type and amounts of effluents that may be released offsite. In addition, this amendment request will not significantly increase individual or cumulative occupational radiation exposures. Therefore, NNECO has determined the proposed changes will not have a significant effect on the quality of the human environment.
Plant Operations Review Committee and Nuclear Safety Assessment Board The Plant Operations Review Committee and Nuclear Safety Assessment Board have reviewed and concurred with these determinations.
Schedule We request. issuance at your earliest convenience, with the amendment to be implemented within 30 days of issuance.
State Notification in accordance with 10CFR50.91(b), a copy of this license amendment request is being provided to the State of Connecticut.
I
~.
t U.S. Nucirr R:gul: tory Commission B16471/Page 3 l
1 l
If you should have any questions on the above, please contact Mr. Ravi Joshi at (860) i 440-2080.
1 Very truly yours l
NORTHEAST NUCLEAR ENERGY COMPANY l
Millstone Unit No. 2 Recove/
M. L. Bowling ry Officer Subscribed and sworn to before me this a0 day of Af 1997 j
1 1%cdww \\CMdued l
u l
Date Commission Expires: id30lM DONNA LYNNE MLLIANS
~
l Attachments (5)
NOTARY PU8 tlc C#
STATE OF CONNFPUCUT cc:
H. J. Miller, Region l Administrator D. G. Mcdonald, Jr., NRC Project Manager, Millstone Unit No. 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit No. 2 Dr. W. D. Travers, Director, Special Projects W. D. Lanning, Director, Millstone Assessment Team Kevin T. A. McCarthy, Director l
Bureau of Air Management Monitoring and Radiation Division l
Department of Environmental Protection l
79 Elm Street Hartford, CT 06106-5127 1
.~
1 Docket No. 50-336 B16471 i
Millstone Nuclear Power Station, Unit No. 2 i
Proposed Revision to Technical Specifications Containment Isolation Valves Discussion of Proposed Changes l
l l
l May 1997 l
_. _ _ _ _. _ _ _ _ _ _. _.. _ _ _..~___
c Li U.S. Nuclecr Regul tory Commission B16471/ Attachment 1/Page 1
\\
Proposed Revision to Technical Specifications Containtnent isolation Valves Discussion of Proposed Changes introduction l
Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO), in a letter dated January 22,1996, proposed to amend Operating License DPR-65 by changing the Technical Specifications of Millstone Unit No. 2. The proposed changes modified l
Definition 1.8 " Containment Integrity," Technical Specification 3.6.1.1 " Containment integrity," Technical Specification 3.6.3.1 " Containment isolation Valves," and the Bases for Technical Specification 3.6.3.1 " Containment isolation Valves." The proposed change relocated the list of containment isolation valves, Table 3.6-2, from the Technical Specifications to the Technical Requirements Manual (TRM), in accordance with NRC Generic Letter (GL) 91-08', which provides guidance for removal of component lists from the Technical Specifications.
l A review of the original submittal revealed deviations from the guidance contained in GL 91-08. These deviations are not necessary. Therefore, NNECO is proposing to rectify this by revising the original submittal, dated January 22,1996, with this new l
submittal. This new submittal is consistent with the recommendations contained in GL 91-08.
The original submittal also modified Surveillance Requirement (SR) 4.6.1.1.a to allow the position of valves, blind flanges, and deactivated automatic containment isolation valves located inside containment to be checked during each cold shutdown, but not more often than once per 92 days. The current requirement is to verify the position of valves, blind flanges, and deactivated automatic containment isolation valves located inside containment once per 31 days. This submittal requests the same change, however the wording has been modified.
l This submittal also expands Bases Section 3/4.6.3, " Containment Isolation Valves," to address the appropriate administrative controls for operation of containment isolation valves in Modes 1 through 4. Since this was not contained in the original submittal, the original Safety Assessment and Significant Hazards Consideration have been modified to address this addition.
l l
J.G. Partlow letter to All Holders of Operating Licenses or Construction Permits for Nuclear Power Reactors, " Removal of Component Lists from Technical Specifications (Generic Letter 91-08)," dated May 6,1991.
U.S. Nuclear Regulatory Commission B16471/ Attachment 1/Page 2 Descriotion of Proposed Chanaes The proposed changes will:
1.
Relocate the list of containment isolation valves contained in Table 3.6-2 of i
Technical Specification 3.6.3.1, " Containment isolation Valves," to the TRM in accordance with the guidance contained in GL 91-08.
The TRM list will contain all of the containment isolation valves specified in the Millstone Unit No. 2 FSAR, of which Technical Specification Table 3.6-2 is a subset.
The penetrations and associated valves addressed by the current i
Technical Specification definition of containment integrity, and associated LCO (3.6.1.1, " Containment Integrity"), are the some as contained in the Millstone Unit No. 2 FSAR.
Therefore, no additiond valves will be identified as containment isolation valves, and no additional vaivas will require surveillance, as a result of this change.
2.
Provide an exception to the surveillance requirement that verifies the position of containment isolation valves, located inside containment, that are locked, sealed, or otherwise secured in the closed position when the unit is in Mode 4 or above.
3.
Expand the Bases of the affected' Technical Specification to specify the l
appropriate administrative controls to allow manual containment isolation valves to be opened.
Chanaes l
1.
The Containment integrity definition (1.8) has been modified to include a reference to the administrative controls of Technical Specification 3.6.3.1. This will allow closed manual valves to be opened, provided the appropriate administrative controls are established. A discussion of the appropriate administrative controls has been added to the Bases for Technical Specification 3.6.3.1. This change is consistent with GL 91-08.
2.
A note (**) has been added to SR 4.6.1.1.a. This note applies to valves, blind flanges, and deactivated automatic containment isolation valves located inside containment. The note changes the surveillance. requirements for these components. Currently, the position of these components is verified every 31 days when the plant is in Modes 1 through 4. The proposed change will require position verification prior to entering Mode 4 from Mode 5, if not performed within j
the previous 92 days. The note is being added so that containment entry, at power, for position verification, will not be required. The wording of this note has j
been ' taken from NUREG-0212 Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors Revision 2 (LCO 3.6.1.1) 3 l
U.S. Nucl:ar Regulatory Commission B16471/ Attachment 1/Page 3 ind NUREG-1432 Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors Revision 1 (LCO 3.6.3).
3.
SR 4.6.1.1.a. has been modified to include a reference to the administrative controls of Technical Specification 3.6.3.1. This will allow closed manual valves to be opened, provided the appropriate administrative controls are established.
A discussion of the appropriate administrative controls has been added to the Bases for Technical Specification 3.6.3.1. This change is consistent with GL 91-08.
4.
The words "specified in Table 3.6-2" have been removed from LCO 3.6.3.1 and associated Action statement, SR 4.6.3.1.1, and SR 4.6.3.1.2. Table 3.6-2, which contained the list of containment isolation valves, has been deleted. The containment isolation valves contained in this table have been relocated to the TRM. The word "The" has been replaced with "Each" and " valves" has been changed to " valve" in LCO 3.6.3.1 and SR 4.6.3.1.1 as a result of the deletion of 1
this table. Also, the word "as" has been removed from SR 4.6.3.1.1 as a result of the deletion of this table. This change is consistent with GL 91-08.
5.
A note (*) has been added to LCO 3.6.3.1. This note allows locked or sealed closed containment isolation valves to be opened on an intermittent basis, in Modes 1 through 4, provided the appropriate administrative controls have been established. A discussion of the appropriate administrative controls has been added to the Bases for Technical Specification 3.6.3.1. This change is consistent with GL 91-08.
6.
SR 4.6.3.1.1.b has been changed by adding a comma after maintenance. This change is editorial. It does not change the intent of this surveillance requirement.
7.
Table 3.6-2 has been deleted and "THIS PAGE INTENTIONALLY LEFT BLANK" added to the blank pages. The containment isolation valves that were listed in Table 3.6-2 have been relocated to the TRM. This change is consistent with GL 91-08.
8.
A statement about where the list of containment isolation valves is located and a discussion of the appropriate administrative controls has been added to the Bases for Technical Specification 3.6.3.1. The discussion includes information that is consistent with GL 91-08. Additional information has been added to address specific requirements for certain Millstone Unit No. 2 valves.
Desian Basis and Licensina Basis l
The containment isolation valves are used to close all fluid (liquid and gas) penetrations not required for operation of the engineered safety feature systems, to prevent the leakage of radioactive materials to the outside environment. The fluid i
i l
l t
.. -.. - - ~. ~. - - - -. - - - - -. - - -
-.. ~. - - _.. _
U.S. Nuclear Ragul tory Commission l
l B16471/ Attachment 1/Page 4 penetrations which may require isolation after an accident are categorized as Type P, O, or N. The penetration types are listed with the containment isolation valves in the Millstone Unit No. 2 FSAR.
Type P penetrations are lines that connect to the reactor coolant pressure boundary (Criterion 55 of 10CFR50, Appendix A). These lines are provided with two containment isolation valves, one inside containment, and one outside containment.
Type O penetrations are lines that are open to the containment internal atmosphere j
(Criterion 56 of 10CFR50, Appendix A). These lines are provided with two containment isolation valves, one inside containment, and one outside containment.
Type N penetrations are lines that neither connect to the reactor coolant pressure boundary, nor are open to the containment internal atmosphere, but do form a closed system within the containment structure (Criterion 57 of 10CFR50, Appendix A). These lines are provided with single containment isolation valves outside containment. These valves are either remotely operated or locked closed manual valves.
Administrative Controls Locked or sealed closed containment isolation valves may be opened on an intermittent basis, provided appropriate administrative controls are established. The position of the NRC, concerning acceptable administrative controls, is contained in GL 91-08, and includes the following considerations:
(1) stationing an operator, who is in constant communication with the control room, at the valve controls; (2) instructing this operator to close these valves in an accident situation; and (3) assuring that environmental conditions will not preclude access to close the valve and that this action will prevent the release of radioactivity outside the containment.
A review of the containment isolation valves, that do not receive an automatic containment isolation signal and that may require operation in Modes 1 through 4, has been performed.
This review identified the containment isolation valves, whose operation in Modes 1 through 4, will require the establishment of appropriate administrative controls in accordance with the guidance added by this change (consistent with the guidance contained in GL 91-08).
l All of the containment isolation valves identified by this review, that will require local i
operation, are located outside of containment.
In addition, these valves are all l
associated with Type N penetrations (except 2-SI-709, which is addressed separately).
This type of penetration does not connect directly to either the reactor coolant pressure i
U.S. Nucinr Reguintory Commission B16471/ Attachment 1/Page 5 boundary, or to the containment intamal atmosphere. For environmental conditions to preclude access to close any of the valves immediately following an accident, a failure of the containment structure or the penetration piping in the vicinity of the valve will be required. In addition, for the closure of the valve to be neG* : ry to stop the release of
~
radioactivity outside of containment (as a result of tt iccident), a failure of the associated closed system (e.g., steam generator tube '.&re) within containment will be required. Therefore, it is extremely unlikely conditions would develop, immediately after an accident, that would prevent the dedicated operator from closing any of these valves.
The appropriate administrative controls for each of these valves is discussed below.
1.
2-SI-463 This is the safety injection tank (SIT) recirculation header stop valve.
It is opened to fill or drain the SITS and for Shutdown Cooling System (SDC) boron equalization. This valve is located outside of containment, and is associated l
with a Type N penetration.
When 2-SI-463 is open, a dedicated operator, in continuous communication with the control room, is required.
2.
2-FW-15A and 2-FW-158 These steam generator chemical addition valves are opened to add chemicals to the steam generators using the Auxiliary Feedwater System (AFW). These j
valves are located outside of containment, and are associated with Type N i
When either 2-FW-15A or 2-FW-158 is opened, a dedicated operator, in continuous communication with the control room, is required.
3.
2-MS-458 and 2-MS-459 These valves, which bypass around the main steam supplies to the turbine driven auxiliary feedwater pump (2-MS-201 and 2-MS-202), are opened to drain water from the steam supply lines. These valves are located outside of containment, and are associated with Type N penetrations.
When either 2-MS-458 or 2-MS-459 is opened, a dedicated operator, in cniinuous communication with the control room, is required.
i U.S. Nucl:tr Regulatory Commission B16471/ Attachment 1/Page 6 4.
2-MS-190A and 190B Operation of these atmospheric steam dump valves is normally. performed by remote operation from the control room. However, local operation may be necessary.
These valves are located outside of containment, and are associated with Type N penetrations.
When either 2-MS-190A and 1908 is operated locally, a dedicated operator, in constant communication with the control room, is required.
The requirement for a dedicated operator, in constant communication with the control room, can be applied to local operation of remotely operated valves associated with Type N fluid penetrations. Local operation of these valves with a dedicated operator is equivalent to the operation of other manual (locked or sealed closed) containment isolation valves with a dedicated operator.
5.
2-SI-045 l
This is the nitrogen header drain valve.
It is opened to depressurize the containment side of the nitrogen supply header stop valve,2-SI-312. This valve is located outside of containment, and is associated with a Type N penetration.
When 2-SI-045 is opened, a dedicated operator, in continuous communication with the control room, is required.
6.
2-SA-19 This is the containment station air header isolation.
It is opened to supply station air to containment. This valve is located outside of containment, and is associated with a Type N penetration.
When 2-SA-19 is opened, a dedicated operator, in continuous communication with the control room, is required.
I 7.
2-IA-566 This is the backup air supply master stop. It is opened to supply backup air to 2-CH-517, 2-CH-518, 2-CH-519, 2-EB-88, and 2-EB-89. This valve is located
.outside of containment, and is associated with a Type N penetration.
When 2-IA-566 is opened, a dedicated operator, in continuous communication with the control room, is required.
U.S. Nucl:rr Rtgulatory Commission B16471/ Attachment 1/Page 7 8.
2-GR-63 This is the containment waste gas header test connection isolation valve. It is opened to sample the primary drain tank for oxygen and nitrogen. This valve is located outside of containment, and is associated with a Type N penetration.
When 2-GR-63 is opened, a dedicated q.'erator, in continuous communication with the control room, is required.
j Three containment isolation valves, associated with Type P penetrations, were also identified by this review. This type of penetration connects directly to the reactor coolant pressure boundary. Two of these valves are located inside containment, and one is outside of containment. Each of these valves is discussed below.
l 1,
2-SI-651 i
This valve is the inside containment isolation valve for the containmant-penetration associated with the inlet to the Shutdown Cooling (SDC) System. It is normally operated from the control room. It does not receive an automatic containment isolation closure signal, but is interlocked to prevent opening if Reactor Coolant System (RCS) pressure is greater than approximately 275 psia.
When this valve is opened from the control room, either one of the two required licensed (Reactor Operator) control room operators can be credited as the dedicated operator required for administrative control. it is not necessary to use l
a separate dedicated operator.
1 i
This discussion of the appropriate administrative control for operation of 2-SI-651 is only valid during SDC operation because RCS pressure and temperature are significantly below normal operating pressure and temperature (the RCS is administratively required to be < 300 F and < 265 psia before SDC flow is initiated).
2.
2-SI-709 This valve is the outside containment isolation valve for the containment penetration associated with the inlet to the SDC System. It is opened locally.
i I
t i
U.S. Nucir:r R:gulatory Commission B16471/ Attachment 1/Page 8 After this valve is opened locally, a separate dedicated operator is not required to remain at the valve. 2-SI-709 is opened before 2-SI-651. Therefore, opening 2-SI-709 will not establish a connection between the RCS and the SDC System.
Opening 2-SI651 will connect the RCS and SDC System. If a problem then develops,2-SI-651 can be closed from the control room. in addition, another valve, 2-SI-652, is located in the SDC inlet flowpath upstream of 2-SI-651 which could be closed to isolate this penetration. Therefore, it is not necessary to station a dedicated operator at 2-SI-709 to ensure the penetration can be isolated if an accident occurs.
i This discussion of the appropriate administrative control for operation of 2-SI-709 is only valid during SDC operation because RCS pressure and 1
temperature are significantly below normal operating pressure and temperature (the RCS is administratively required to be < 300 F and < 265 psia before SDC
]
flow is initiated).
I 3.
2-CH-517 This is the pressurizer auxiliary spray valve. It can be used as an alternate method to decrease pressurizer pressure, or for boron precipitation control following a loss of coolant accident.
When this valve is opened from the control room, either one of the two required licensed (Reactor Operator) control room operators can be credited as the dedicated operator required for administrative control. It is not necessary to use a separate dedicated operator.
This is acceptable because the fluid that passes through this valve will be collected in the Preasurizer (reverse flow from the Pressurizer to the charging system is pmvented by check valve 2-CH-431), and the penetration associated with 2. ' 4-5) 7 is open during accident conditions to allow flow from the charging pumps. Also, this valve is normally operated imm the control room, under the supervision of the licensed control room ope,ators, in accordance with plant procedures.
Safety Assessment The proposed changes will:
1.
Relu: ate the containment isolation valve (CIV) list, Table 3.6-2, from the Technical Specifications (LCO 3.6.3.1) to the TRM in accordance with GL 91-08.
2.
Modify SR 4.6.1.1.a to allow the valves, blind flanges and deactivated automatic valves located inside the containment that are locked, sealed, or otherwise secured in the closed position to be verified closed prior to entering Mode 4 from
4 U.S. Nucirr R:gulatory Commission B16471/ Attachment 1/Page 9 Mode 5, if not performed within the previous 92 days, instead of the current 31 day requirement.
3.
Expand the Bases of the affected Technical Specification to specify the appropriate administrative controls to allow manual containment isolation valves to be opened.
Relocation of the ClV list does not affect the designation of any individual valve as a CIV, nor does it reduce the operability and testing requirements for any valve. Any l
change to the TRM list will require a 10CFR50.59 evaluation.
The 10CFR50.59 l
evaluation will review the three criteria of 10CFR50.59(a)(2) to determine if the proposed change involves an unreviewed safety question. Also, a Plant Operations Review Committee (PORC) review and approval of the proposed change will be required. This process will ensure that a thorough review is performed before changes are made to the CIV list. This proposed change does not diminish any other regulatory or Technical Specification requirement for the valves. Since this change only involves the type of controlled document used to maintain the list of CIVs, and does not affect the operability, maintenance, testing, or use of the valves, there is no adverse effect on plant safety.
The change to SR 4.6.1.1.a will allow the valves, blind flanges and deactivated automatic valves located inside the containment that are locked, sealed, or otherwise secured in the closed position to be verified closed prior to entering Mode 4 from Mode 5, if not performed within the previous 92 days, instead of the current 31 day requirement. This means that the surveillance interval could be as long as the entire operating cycle, depending on whether entry into Mode 5 is required during the cycle.
The change in the surveillance frequency (increase in time from 31 days to not less than 92 days and only prior to entering Mode 4 from Mode 5) recognizes that these valves are operated under administrative controls and the probability of misalignment is low. This proposed change to SR 4.6.1.1.a is consistent with NUREG-1432 Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors Revision 1 (LCO 3.6.3).
The additional information added to the Bases will specify the appropriate administrative controls necessary to allow manual containment isolation valves to be opened. These administrative controls were developed after consideration of the three items contained in GL 91-08, and repeated in the proposed addition to Bases Section 3/4.6.3.
l The proposed changes do not have any adverse impact on the design basis accidents l
previously evaluated. Therefore, there is no significant impact on the public health and j
safety.
i i
_m_.
l 1
\\
Docket No. 50-336 l
i B16471 i
l l
i l
l l
Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Containmeniisolation Valves Significant Hazards Consideration t
l l
l l
l i
May 1997 l
U.S. Nuclur Reguintory Commission B16471/Aticchment 2/Page 1 Proposed Revision to Technical Specifications Containment isolation Valves Significant Hazards Consideration Sionificant Hazards Consideration In accordance with 10CFR50.92, Northeast Nuclear Energy Company (NNECO) has reviewed the proposed changes and has concluded that they do not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve an SHC because the changes do not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to relocate the containment isolation valve (CIV) list will not result in any hardware or equipment operating changes. The proposed change is based on Generic Letter (GL) 91-08 and merely relocates the CIV table and removes all references to the table. The relocation of the CIV table from the Technical Specifications does not affect the operability requirements of any of the listed valves. Technical Specifications will still continue to require the CIVs to be operable. The LCO and surveillance requirements for the valves will remain in Technical Specifications.
The CIV table will be relocated to the Millstone Unit No. 2 Technical Requirements Manual (TRM), which is controlled in accordance with 10CFR50.59.
This change does not alter the design, function, or operation of the valves involved. Thus, there is no significant affect on the probability or consequences of any previously evaluated accident.
The change to Surveillance Requirement (SR) 4.6.1.1.a will allow the valves, blind flanges and deactivated automatic valves located inside the containment that are locked, sealed, or otherwise secured in the closed position to be verified closed prior to entering Mode 4 from Mode 5, if not performed within the previous 92 days, instead of the current 31 day requirement. This means that the surveillance interval could be as long as the entire operating cycle, depending on whether entry into Mode 5 is required during the cycle. The change in the surveillance frequency (increase in time from 31 days to not less than 92 days and only prior to entering Mode 4 from Mode 5) recognizes that these valves are operated under administrative controls and the probability of misalignment is low. This provides adequate assurance that the containment function assumed in the accident analysis will be maintained. Therefore, there is no significant affect on the probability or consequences of any previously evaluated accident. This proposed change to SR 4.6.1.1.a is consistent with NUREG-1432 Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors Revision 1 (SR 3.6.3.4).
U.S. Nucle:r Rtgulttory Ccmtrission l
B16471/ Attachment 2/Page 2 The information added to the Bases will provide additional guidance to ensure the plant is operated correctly. This information will not result in any new approaches to plant operation. Therefore, there is no significant affect on the probability or consequences of any previously evaluated accident.
These proposed changes do not alter the design, function, or operation of the valves involved. Therefore, there is no significant increase in the probability or i
consequence of an accident previously evaluated.
l 2.
Create the possibility of a new or different kind of accident from any accident l_
previously evaluated.
l
[
The change to relocate the CIV list trom the Technical Specifications to the TFM l
will not impose any different operational or surveillance requirements, nor wiii the change remove any such requirements.
Adequate control will be j
maintained. Furthermore, as stated above, the proposed change does not alter the design, function, or operation of the valves involved, and therefore does not l
create the possibility of a new or different kind of accident from any accident previously evaluated.
l I
The change to SR 4.6.1.1.a reduces the surveillance frequency for valves, blind flanget and deactivated automatic valves located inside the containment. It does not alter the design, function, or operation of the valves. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The information added to the Bases will provide additional guidance to ensure the plant is operated correctly. This information does not alter the design, function, or operation of the valves involved. Therefore, it does not create the possibility of e new or different kind of accident from any accident previously evaluated.
3.
Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since they have no impact on any safety analysis assumption.
The proposed changes do not decrease the scope of equipment currently required to be operable or subject to surveillance testing, nor do the proposed changes affect any instrument setpoints or equipment safety functions.
l The effectiveness of Technical Specifications will be maintained since the change will not alter function or operability requirements for any CIV.
In addition, the relocation of the valve list is consistent with the guidance provided j
in GL 91-08, and the change to the surveillance interval is consistent with NUREG-0212 Standard Technical Specifications for Combustion Engineering l
U.S. Nucl=r R:guintory Commission B16471/ Attachment 2/Page 3 Pressurized Water Rwiors Revision 2 (LCO 3.6.1.1) and NUREG-1432 Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors Revision 1 (LCO 3.6.3).
The information added to the Bases is consistent with the guidance provided in GL 91-08 and NUREG-1432 Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors Revision 1.
The intent of the Technical Specifications will be met since this information will not result in any new approaches to plant operation.
Therefore, there is no significant reduction in a margin of safety.
The NRC has provided guidance conceming the application of standards in 10CFR50.92 by providing certain examples (March 6, 1986, 51 FR 7751) of amendments that are considered not likely to involve an SHC. The changes propo sed herein are not enveloped by a specific exampie. However, the proposed changes do not alter the design, function, or operation of the equipment involved. Therafore, NNECO has concluded that the proposed changes do not involve an SHC.
1
l e
Docket No. 50-336 1
B16471 j
I J
l 4
i 4
i Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Containment Isolation Valves New Marked Up Pages May 1997