ML20202B823
| ML20202B823 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/19/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20202B822 | List: |
| References | |
| GL-91-08, GL-91-8, NUDOCS 9712030188 | |
| Download: ML20202B823 (6) | |
Text
.
A p* ** 00 y
t UNITED STATES g
]
NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 30685 4001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 210 TO FACILITY OPERATING LICENSE NO. DPR-65 NORTHEAST NUCLEAR ENERGY C00 ANY THE CONNECTICUT LIGHT AND POWER COMPANY AND WESTERN MASSACHUSETTS ELECTRIC COMPANY MILLSTONE NUCLEAR POWER STATION. UNIT 2 DOCKET NO. 50-336
1.0 INTRODUCTION
By letter dated May 20, 1997, as supplemented on September 23, 1997, the Northeast Nuclear Energy Company, et al. (the licensee) submitted a request for changes to the Millstone Nuclear Power Station, Unit 2, Technical Specifications (TSs). The changes would relocate the containment isolation valve (CIV) list from the TSs to the Technical Requirements Manual (TRM) in accordance with Generic Letter (GL) 91-08, " Removal of Component Lists from the Technical Specifications." The request would also change the surveillance requirement for valves, blind flinges, and deactivated automatic valves l
located inside the containment tiat are locked, sealed, or otherwise secured in the closed position from once avery 31 days to during each cold shutdown, but no more than once per 92 days.
The TS Bases would be changed to reflect i
the relocation of the CIV list from the TSs to the TRM and includes the l
administrative controls for CIV o)eration in Modes 1 through 4.
The i
September 23, 1997, letter provided clarification relating to two of the CIVs I
that can be secured from the control room. The additional information did not affect the initial proposed no significant hazards consideration determination.
The affected TS Sections are:
TSs 1.8.1.b, 4.6.1.1.a. 3.6.3.1, 4.6.3.1.1, 4.6.3.1.l.b 4.6.3.1.2, Table 3.6-2, and TS Bases 3/4.6.3.
A license condition has been added to paragraph 2.C. of the Operating License.
2.0 BACKGRQM@
t l
Section 182a of the Atomic Energy Act of 1954, as amended (the "Act") requires applicants for nuclear power plant operating licenses to include TSs as part of the license.
The Commission's regulatory requirements related to the content of-TSs are set forth in 10 CFR 50.36. That regulation requires that j
the TSs include items in five specific categories, including (1) safety 9712030188 971119 l
PDR ADOCK 05000336 i
P PDR l
' limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.
However, the regulation does not specify the particular requirements to be included in a plant's TSs.
The Commission has provided guidance for the contents of TSs in its " Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (" Final Policy Statement"), 56 FR 39132 (July 22,1993), in which the Commission indicated that compliance with the Final Policy Statement satisfies Section 182a of the Act.
In particular, the Commission indicated that certain items could be relocated from the TSs to licensee-controlled documents, consistent with the standard enunciated in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979).
In that case, the Atomic Safety and Licensing Appeal Board indicated that " technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety."
Consistent with this approach, the Final Policy Statement identified four criteria to be used in determining whether a particular matter is required to be included in the TSs.
These criteria were subsequently incorporated into the regulations by an amendment tc 10 CFR 50.36, 60 FR 36953 (July 19, 1995).
The criteria incorporated into the rule are as follows:
(1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (3) a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; and (4) a structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.
As a result, existing limiting conditions for operation requirements, which fall within or satisfy any of the criteria, must be retained in the TSs, while those TS requirements, which do not fall within or satisfy these criteria, may be relocated to other licensee-controlled documents.
3.0 EVALUATION In relation to the first portion of the nquest, the licensee has proposed that the list of CIVs in TS Table 3.6-E be removed from the TSs and relocated to the TRM.
Any table notations must be relocated to other TSs and references to the table must be deleted when a TS table is removed.
The licensee indicated that the TRM list will contain all of the CIV's specified in the Updated Final Safety Analysis Report (UFSAR) of which the CIVs listed in TS Table 3.6-2 are a subset.
UFSAR Section 12-9, " Technical Requirements Manual," indicates that the technical requirements portion l
l
. (Section 2.0 of the TRM) is incorporated by reference.
It further indicates that changes to the TFA require a 10 CFR 50.59 safety evaluation.
TS Table 3.6-2 has only one notation that allows the manual valves to be opened on an intermittent basis under administrative control. The licensee has proposed that this provision requiring administrative controls be relocated as a footnote to TS 3.6.3.1 and that TS 4.6.1.1.a include a reference to TS 3.6.3.1.
TS 1.8.1.b, which defines containment integrity, also is modified to reference the administrative controls referred to in TS 3.6.3.1.
The licensee also proposes that the references to TS Table 3.6-2 in TSs 3.6.3.1, 4.6.3.1.1, and 4.6.3.1.2 be deleted.
These proposals are consistent with the guidance in GL 91-08 and are, therefore, acceptable.
TS Table 3.6-2 also specifies the maximum isolation times for the automatic CIVs, which the licensee proposes to remove and will be in the TRM.
TS 4.0.5 requires inservice testing (IST), which includes valve stroke times for a broad class of valves including the automatic CIVs.
Thus, the IST requirement to verify that the stroke times are within the required limits remains in TS 4.0.5 and the removal of the stroke times specified in the TS table is consistent with the guidance in GL 91-08 and is, therefore, acceptable.
In GL-91-08, the NRC staff restated li.s position on considerations that constitute an acceptable administrative control for opening normally closed CIVs. The guidance in the GL indicated that the considerations should be stated in the TS Bases.
The considerations included: (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 valves and that this action will prevent the release of radioactivity outside of the containment.
The licensee proposes to modify TS Bases Section 3/4.6.3 to identify the appropriate administrative controls for the CIVs that are exoected to be opened during operation in Modes 1 through 4.
Eleven CIVs, which will require local operation, are located outside of the i
containment and are Type N penetrations. Type N penetrations are lines that I
neither connect to the reactor coolant pressure (RCS) boundary nor are open to i
the containment internal atmosphere, but do form a closed system within the containmant structure. When any of these valves are opened, administrative controls require that a dedicated operator, in continuous communication with i
l the control room, be stationed at the valve.
Three CIV Type P valves, which are valves that connect directly to the RCS, are expected to require limited opening during operation in Modes 1 through 4.
Two of the valves are inside contairaent and one is outside containment. The two valves located inside containment have remote manual operation capability from the control room and receive no automatic signal.
Two of the CIVs are associated with the shutdown cooling (SDC) system and are only opened under administrative controls during SDC when the RCS temperature l
l
. is less than 260'f and the pressure is less than 265 psig.
This evolution involves a relative short time period of about 3 or 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during which plant conditions are closely monitored until Mode 5 is reached when containment isolation is no longer required.
One of the CIVs is inside containment, is normally operated from the control room, has no automatic isolation signal, and is interlocked to prevent opening when the RCS is greater than 275 psia.
The other CIV in the SDC system is located outside containment and is opened locally. This valve is opened before the other valve is opened from the control room and does not establish a aa'th between the RCS and SDC system. Only when the first valve is suasequently opened from the control room is a path between the RCS and SDC system established through the two valves. The licensee proposes not to station a dedicated operator at this location.
Normally each of the CIVs are required to be capable of being manually closed independently of each other.
However, the licensee indicates that there is another normally open valve located inside of containment that could also be closed from the control room to isolate the penetration should the valve located outside of containment be open for SDC.
Considering the interlocks on the SDC system valves, the limited time periods they would be operated, the relatively low RCS temperature and pressure when the CIV located outside of containment is expected to be opened and the other valve that can be closed, the NRC staff has determined that stationing a dedicated operator at the manual SDC system CIV located outside of the containment is not needed.
The administrative procedures credit one of the two required licensed control room operators, although the operator will not be dedicated, as responsible for manually shutting the CIV from the control room when containment isolation is required.
Thus, the process for closure of the CIV witn remote manual operation capability from the control room is essentially the same as for the CIVs located outside of containment in that the control room operators make the decision to close the CIVs, do not have to consider the time required to direct another operator to manually close a CIV in a remote location or be concerned about a potentially adverse environment at a remote location.
If a problem then develops, the first CIV can be closed remotely from the control room to provide isolation between the RCS and the SDC system.
The other CIV is located inside containment, normally operated from the l
control room, and is in the pressurizer auxiliary spray system.
This CIV is opened as an alternative method to decrease pressurizer pressure, or for boron precipitation control following a loss-of-coolant accident.
Any fluid that passes through this CIV is collected in the pressurizer and the CIV is open during accident conditions to allow flow to the charging pumps.
Backflow to the charging system is prevented by a check valve. As in the SDC system CIVs, a control room operator is credited for closing the CIV when required.
Da the basis of the discussion above, the NRC staff has determined that the CIVs associated with SDC and the pressurizer auxiliary spray system have adequate administrative controls to assure that containment isolation can be achieved when required.
l
4 All the CIVs, which require local operation, are located outside containment and are associated with Type N penetrations (except the one SDC system CIV that would not require reclosing as discussed above).
The licensee indicated that the CIVs would be accessible immediately following an accident to allow them to be closed.
To have an adverse environment in the vicinity of the CIVs, two failures would have to occur.
Ti;e failure of the containment structure or piping near a CIV and an additional failure in the system associated with the CIV since these CIVs neither connect to the RCS nor are open to the containment atmosphere.
Thus, the dedicated operator would be able to close any of the CIVs associated with the Type N penetrations.
The. licensee has adequately addressed the guidance provided in GL 91-08 for establishing administrative controls, as discussed above, and proposes changes to TS Bases 3/4.6.3 to reflect the administrative controls, therefore, we find the proposed changes acceptable.
In summary, these requirements, as previously discussed, are not required to be in the TSs under 10 CFR 50.36 or Section 182a of the Act, and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.
Further, they do not fall within any of the four criteria set fourth in the Commission's Final Policy Statement and subsequently incorporated into 10 CFR 50.36.
Additionally, the NRC staff finds that significant regulatory controls exist under 10 CFR 50.59. Accordingly, the NRC staff has concluded that these requirements may be relocated to the TRM.
The second portion of the request would change the surveillance requirement in TS 4.6.1.1.a for valves (manual), blind flanges, and deactivated automatic CIVs located inside containment that are locked, scaled, or otherwise secured in the closed position from the current requirement of once every 31 days to during each cold shutdown, if they had not been performed within the previous 92 days. The current surveillance requirement in TS 4.6.1.1.a of once per 31 days is applicable for Modes 1, 2, 3, and 4.
l The licensee notes that the current requirement in NUREG-1432, " Standard l
Technical Specifications for Combustion Engineering Plants," f aoicates that l
the surveillances for the valves located inside containment be performed prior l
to entering Mode 4 from Mode 5 if they had not been performed within the i
previous 92 days. The Bases Section of NUREG-1432 indicates that the 31-day l
surveillance interval chosen for those CIVs outside containment was based on engineering judgement and the relative ease that visual verification can be made.
For those CIVs inside containment, the surveillance interval, prior to entering Mode 4 from Mode 5, if not performed within the previous 92 days, is l
appropriate since these CIVs are operated under administrative controls and l
the probability of their misalignment is luw.
It should also be noted that l
access to the CIVs located inside containment would be limited during l
operation in Modes 1, 2, 3, and 4.
Therefore, since the manual CIVs, blind flanges or deactivated CIVs (1) are located inside containment, (2) are locked, sealed, or otherwise secured in f
the closed position, (3) have limited access during power operation, and (4) are consistent with our current surveillance interval requirements, the NRC l
staff has determined that the proposed surveillance interval is acceptable.
I
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Connecticut State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The amendment also relates to changes in recordkeeping, reporting, or administrative procedures or requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation-exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 33128 dated June 18,1997).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (10).
Pursuant to 10 CFR Sl.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the healt5i and safety of the public.
Principal Contributor:
D. Mcdonald Date: Novanber 19, 1997