ML20203M432
| ML20203M432 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/03/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203M429 | List: |
| References | |
| NUDOCS 9803060396 | |
| Download: ML20203M432 (12) | |
Text
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON. D.C. 30MH001
'g +... + g' SAFETY EVALUATION 3Y THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 120 TO FACilllY OPERATING LICENSE NO. DPR 80 AND AMENDMENT NO. 118 TO FACILITY OPERATING LICENSE NO. DPR 82 PACIFIC GAS AND ELECTRIC COMPANY DIAPLO CANYON NUCLEAR POWER PLANT. UNITS 1 AND 2 DOCKET NOS 50 275 AND 50 323
1.0 INTRODUCTION
By application dated October 4, 1995, as supplemented by letters dated August 20. 1996, and June 2.1997. Pacific Gas and Electric Company (PG&E or the licensee) requested changes to the Technical Specifications (Appendix A to l
Facility Operating License Nos. DPR 80 and DPR-82) for the Diablo Canyon Nuclear Power Plant. Units 1 and 2.
i The proposed changes would relocate the i
requirements in ten subsections of the Technical Specifications (TSS) to licensee-controlled documents The July 17, 1996. August 20. 1996, and June 2. 1997, sup)lemental submittals provided additional clarifying information that did not clange the portion of the initial n^ significant hazards consideration determination that was published in the Federal Reaister on November 27, 1995 (60 FR 58404).
2.0 BACKGROUND
l Section 182a of the Atomic Energy Act requires that applicants for nuclear power plant operating licenses shall state:
(S)uch technical specifications, including information of the amount.
kind, and source of special nucleer material required, the place of the use the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization...of special nuclear material will be in accord with the common defense and security and will l
provide adequate protection to the health and safety of the public.
Such technical specifications shall be a part of any license issued.
In-10 CFR 50.36, the Commission established its regulatory requirements related to the content of technical specifications.
i l82 188 M 31888373 P
2 On July 19, 1995, the Commission amended the regulations in 10 CFR 50.36 pertaining to Technical Specifications for nuclear power reactc 1 (60 FR 36953).
The summary stated that "the rule codifies criteria for determining the content of technica! specifications" and that "each licensee covered by these regulations may voluntarily use the criteria as a basis to propose the relocation of existing technical specifications that do not meet any of the criteria from the 4cility license to licensee-controlled documents." The licensee has evaluated the requirements currently in ten subsections of the TSs against the four criteria in the revised 10 CFR 50.36 and against the requirements in the Westinghouse Standard TSs. NUREG.1431. Revision 1. dated April 7, 1995, and concluded that the requirements can be relocated to a licensee controlled document.
In the revised 10 CFR 50.36, the Commission set forth four criteria related to the prevention of accidents and the mitigation of accident consequences. The revised regulation stated that a technical specification limiting condition for operation and associated surveillance requirements must be established for any item meeting one or more of the four criteria.
As noted above, items which do not meet one of the four criteria can be relocated to a licensee document controlled under the provisions of 10 CFR 50.59.
The four criteria are as follows:
1.
Installed instrumentation that is used to detect and indicate in the control room.. a signiticant 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 Casign Basis Accident or Transient that either assumes the failure of or presents a chi. '3nge to the integrity of a fission product barrier.
4 A structure, system, or component which operating experience or 3robabilistic safety assessment has shown to be significant to public lealth and safety.
3.0 EVALUATION The licensee states that the TSs which they 3ropose to remove will be relocated to the Diablo Canyon Power Plant (DCPP) Equipment Control Guidelines (EC3s).
The licensee also stated that these ECGS are cor, trolled by DCPP Department-Level Administrative Procedure (DLAP) OPl.DC16. " Control of Plant Equipment Not Required by the Technical Specifications." The content of the relocated TS will not be changed at the time of relocation.
Future changes to the relocated TS will be made under the provisions of 10 CFR 50.59. as required in DCPP procedure OPl.DC16. This is an acceptable licensee-controlled document.
o
3 3.1 Boration Flow Paths - Operating A limiting condition for operation (LCO) in the present DCPP TSs (TS 3.1.2.2) requires.that both a flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the reactor coolant system (RCS) and a flow path from the refueling water storage tank (RWST) via a charging pump to the RCS must be operable in Modes-1, 2, 3, and 4.
The Bases fer this limiting condition for operation (LCO) state that the purpose is to assure negative reactivity control is available during each mode of facility operation.
The boration subsyster of the chemical and volume control system (CVCS) provides the means.to meet one of the functional requirements of the CVCS, i.e., to control the chemical neutron absorber (boron) concentration in the
-RCS and to help control the boron concentration to maintain shutdown margin (SDM).
To accomplish this functional requirement, the boration systems TS require a source of borated water, one or more flow paths to inject this borated water into the RCS, and appropriate charging pumps to provide the necessary charging head.
The boration subsystem is not assumed to operate to mitigate the consequences of a design basis accident:(DBA) or transient.
In the case-of a malfunction of the CVCS, which causes a boron dilution event; the automatic response,'or
,that required by the operator, is to close the appropriate valves-in the reactor makeup system Defore the SDM is lost.
Operation of the boration subsystem is not assumed to mitigate this event. The normal capability to 4
control reactivity with boron is not credited in the accident analysis.
SDM requirements provide sufficient reactivity mar (,n to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences.
The SDM defines the degree of subcriticality that would-be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod assembly of highest worth is fully withdrawn.
During power operation, SDM control is ensured by operating with tne shutdown banks fully withdrawn and the c M rol banks within the limits of LCOs 3.1.3.5 and 3.1,3.6 for rod insertion.
The.boration subsystem is not installed instrumentation that is used to detect or' indicate a significant degradation of the reactor coolant pressure boundary (RCPB); therefore, this TS does'not satisfy Criterion 1.
The boration subsystem TS is not associated with a process variable that is-an initial condition of an event that assumes failure of or challenges the integrity of a fission product. barrier.
For these events, the primary success path for mitigation includes isolating the dilution flowpath.
The subsequent actuation of equipment to establish a boron injection flowpath is intended to regain the required SDM.
This is desirable, but beyond the scope of a primary success path action.
Therefore, the boration subsystem is not a design feature required to be operable to mitigate these events, and this TS does not satisfy Criterion 2.
l i
4 The boration subsystem TS does not apply to a system that is part of the primary success path, and which functions to mitigate a DBA or transient that either assumes tie failure of or presents a challenge to the integrity of a fission product barrier; therefore, the TS does not satisfy Criterion 3.
Operability of the cha ging pumps, the RWST and associated flowpaths is required as part of the emergency core cooling system (ECCS) TS.
The TS for the ECCS will address the requirements for these components.
For the main steamline break (MSLB) event the sequence of events takes the plant to cold shutdown conditions and; therefore. boration of the RCS is necessary.
However, the boration flowpath in this case is required as part of the ECCS function.
The operating flow paths used to inject borated water to maintain SDM have not been shown to be significant to public health and safety by either operating experience or 3robabilistic safety assessment (PSA).
The boration function is modeled in t1e DCPP Probabilistic Risk Assessment (PRA), but only its function in response to anticipated transient without scram (ATWS) events.
The ATWS contribution to core damage is small, less than 1E 6.
Thus. it can be concluded that this TS does not satisfy Criterion 4 This LCO and the associated surveillance requirements (SR) are not included in Section 3.1. " Reactivity Control Systems." of NUREG-1431. Revision 1. since the requirement does not meet one of the four criteria in 10 CFR 50.36.
The licensee's proposal to relocate TS 3.1.2.2 to the DCPP ECGS. a licensee-controlled document. is acceptable.
3.2 Position Indication System - Shutdown TS 3.1.3.3 of the present TSs requires that one digital rod position indicator be operable and capable of determining control rod Josition in Modes 3. 4. and 5 (Hot Standby. Hot Shutdown, and Cold Shutdown wit 1 zero percent rated thermal power).
In its safety evaluation, the licensee discussed why this requirement does not meet any of the four criteria in 10 CFR 50.36.
The licensee proposes to relocate TS 3.1.3.3 to a licensee-controlled document.
The system will be required by TS 3.1.3.2 to be operable in Modes 1 and 2:
i this requirement will remain in the TSs.
Control rod position is used by the operator to verify that the rods are correctly positioned and to verify that the rods are inserted into the core following a reactor trip.
Rod position is also used during a reactor startup.
Operability of the control red position indicators is required to determine control rod positions and; thereby, ensure compliance with the rod alignment and insertion limits.
These rod alignment requirements are applicable during power operation to maintain power distribution limits.
Rod insertion limits are required to maintain SDM during Modes 1 and 2.
The Bases do not address the shutdown condition.
The LCO requires that one position indicator be operable to determine the position of any rod not fully inserted.
Rod position indication may be used during a control rod withdrawal event from shutdown condition, but it is not required to be operable as an initial condition or mitigating signal.
f:
5-
. The position indication system TS is not applicable to installed instrumentation used to detect and indicate in the control room significant
- abnormal degradation of the RCPB. Therefore, the TS does not meet Criterion 1.
The position indication system TS,.for shutdown conditions, is not associated with a process variable, design feature or operating restriction that is an initial condition of a DBA or-transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product
- barrier. Therefore, the TS does not meet Criterion 2.
Finally the position indication system TS does not apply to a structure, system or component (SSC) that is part of the primary success )ath; and which functions to mitigate a DBA or transient that either assumes tie failure of or presents.a challenge to the integrity of a fission product barrier.
Therefore, the TS does not satisfy Criterion 3.
- The control rod position indicating systems have not been shown to be significant to public health and safety by either operational experience or PSA. The system is not modeled in the DCPP PRA but there is no indication that it would be identified as risk significant if it were included in the i
Therefore, this TS does not satisfy Criterion 4.
In NUREG 1431. Revision 1. LC0 3.1.8 only requires that the digital rod position indicator system be operable in Modes 1 and 2.
The relocation of the requirement in TS 3.1.3.3 to a licensee-controlled document is acceptable.
SR 4.1.3.3 for LC0 3.1.3.3 in the present TSs requires that each of the required digital rod position indicators shall be determined to be operable by verifying that the digital rod position indicators agree with the demand position indicators within 12 steas when exercised over the full range of rod-travel at least once each REFUELI4G INTERVAL. An essentially identical surveillance requirement was retained in NUREG-1431, Revision 1, as SR - 3.1.8.1.
The licensee also pro)oses to retain this surveillance requirement in the TSs by relocating it ver]atim to become a surveillance requirement for LCO 3.1.3.2.
The-relocation of this requirement in the TSs is an administrative change and is acceptable.
.3.3 Rod Drop' Time In-the existing TSs. LC0 3.1.3.4 requires that the individual full-length shutdown and control rod drop time be less than 2.7 seconds prior to proceeding to. Mode 1 or 2.
The associated SR requires that the rod drop time of full-length rods be demonstrated through measurement 3rior to reactor criticality (a) for all rods following each removal of tie reactor vessel head. (b) for specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop
-time of those specific rods, and (c) at least once each REFUELING INTERVAL.
Except for the last surveillance (verification each REFUELING INTERVAL), the licensee proposes to convert the exact requirements which are now in LCO 3.1.3.4 and SR 4.1.3.4 into a new SR 4.1.3.1.3 to LC0 3.1.3.1 on moveable
_w
i
-6 control assemblies.
The relocation of these requirements in the TSs is an administrative change and is acceptable.
3.4 Seismic Instrumentation In the existing TSs. LCO 3.3.3.3 and SR 4.3.3.3.1 and 4.3.3.3.2 specify the operability requirement for seismic instrumentation. This instrumentation does not meet the criteria in 10 CFR 50.36 and is not included in NUREG 1431.
Revision 1.
Relocation of this TS and the associated tables to the DCPP ECGS.
a licensee controlled document. is acceptable.
The TS Bases state that the seismic monitoring instruments are to determine the magnitude of a seismic event 50 that the measured response of the plant can be compared to the response used in the design basis and determine if a shutdown is required in accordance with 10 CFR 100. The occurrence of a seismic event would represent a challenge to fission product barriers.
However the ability of the plant to withstand a desio earthquake (DE).
double design earthquake (DDE), and postulated 7.5M Ho w ri earthquake (HE) is a design requirement.
The seismic monitoring instrumentation performs no role in mitigating a seismic event or in achieving a safe shutdown condition after a seismic event has occurred (the seismic monitoring system is separate from the seismic trip system which is unaffected by this TS relocation).
The seismic instrumentation TS is not ap)licable to installed instrumentation that is used to detect and indicate in tie control room a significant abnormal degradation of the RCPB.
The seismic instrumentation TS does not apply to a SSC that is part of the primary success path, and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to ne integrity of a fission product barrier.
The seismic instrumentation is not associated with a process variable, design feature or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The seismic instrumentation has not been shown to be significant to public health and safety by either operational experience or PSA.
Although seismic events contribute 4E-5/ year to core damage the seismic instrumentation is not modeled in the DCPP seismic PRA-because it does not mitigate the effects of a seismic event.
Based on the above, the seismic instrumentation requirements do not satisfy criteria 1. 2. 3 or 4 3.5 Chlorine Detection Systems In the existing TSs. LC0 3.3.3.7 and SR 4.3.3.7 specify the operability requirements for chlorine detection systems.
This instrumentation does not
4 7-meet the criteria in 10 CFR 50.36 and is not included in the new standard TSs.
Relocation of this TS to a licensee controlled document is acceptable, although gaseous chlorine has been removed from the plant and t1e requirement is no longer applicable.
The chlorine detection system instrumentation is used to detect an accidental chlorine release and 1solate the control room atmosphere.
An accidentel chlorine release is not a DBA or transient.
The chlorine detection system instrumentation is not installed instrumentation that is used to detect, and indicate in the contr01 room, a significant abnormal degradation of the reactor coolant pressure boundary.
The chlorine detection system instruo.ntation does not meet Criterion 1.
The chlorine detection system instrumentation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The chlorine detection system instrumentation does not meet Criterion 2.
The chlorine detection system instrumentation is not assumed to function in the safety analysis. The chlorine detection system instrumentation is not a system that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The chlorine detection system instrumentation does not meet Criterion 3.
Since gaseous chlorine has been removed from the plant, the external events PRA no longer considers any risk from chlorine, Thus, it can be concluded that based on Criterion 4. this TS can be relocated.
3.6 Turbine Overspeed Protection Existing TS 3.3.4.1. 4.3.4.1.1 and 4.3.4.1.2 contain the LC0 and surveillance requirements on the turbine overspeed protection system.
At Diablo Canyon, the system consists of four high-pressure turbine control valves, pressure turbine stop valves, four high-six stop valves, and six reheat intercept valves for the low-pressure turbines, along with the control systems. The system is designed to mitigate a potential overspeed event, not only to protect the turbine but also to prevent the generation of potential missiles such as turbine blades.
The turbine overs]eed event is not a DBA.
This event is evaluated to determine the pro] ability of damage to equipment needed for safe shutdown.
The turbine has a favorable orientation from the standpoint of low trajectory missiles: however, the combination of overspeed probability with high trajectory strike probability must meet the NRC's requirements for overall probability, i.e., less than lE-5 per year.
l
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8-The turbine overspeed protection system is not ap)licable to installed instrumentation used to detect, and indicate in tie control room, a significant abnormal: degradation of the RCPB; therefore, this instrumentation does not satisfy Criterion 1 (turbine overspeed protection is not associated with the RCPB).
The turbine overspeed protection system is not associated with a process variable, design feature or operating restriction that is an initial condition of any DBA or transient analysis. Thus, this instrumentation does not satisfy Criterion 2.
The turbine overspeed protection system is not assumed to function in the safety analysis.
It does not apply to any SSC that is a part of the primary success )ath, and which functions or actuates to mitigate a DBA or transient that eitler assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Thus, this instrumentation does not satisfy Criterion 3.
The turbine overspeed protection has not been shown to be significant to public health and safety by Either operational experience or PSA. Turbine overspeed protection was analyzed and addressed in the DCPP external events
-PRA and was concluded to only contribute a small fraction of the total core damage frequency.
Thus. this TS does not satisfy Criterion 4.
The system does not satisfy any of the four criteria in 10 CFR 50.36 and is not included in NUREG-1431. Revision 1, The relocation of the requirements in-TS 3.3.4.1. 4.3.4.1.1 and 4.3.4.1.2 to a licensee-controlled document, specifically the DCPP ECGS.- is acceptable.
3.7 Containment Leakage Section 3.6.1.2 of the existing TSs identifies the allowable leakage rates for the containment structure to be in'accordance with the. containment leakage rate testing program. The associated surveillance requirement. 4.6.1.2, describes how the test schedule and containment leakage rates shall be determined in conformance with-the criteria s;acified in the containment leakage rate testing program.
The licensee proposes to relocate all of Section 3.6.1.2 to a licensee-controlled document and to add a new surveillance requiremeat to TS 3.6.1.1 (as 4.6.1.lc) on containment integrity. The LCO in existing TS 3.6.1.2 specifies that the leakage rate shall be limited in accordance with the containment leakage rate testing program.
With the relocation of the definitions to the Bases and the containment leakage rate testing program, the relocation of TS 3.6.1.2 to a licensee-controlled document is acceptable as discussed below.
This specification is not apalicable to installed instrumentation that is used to detect. and indicate in t1e-control room, a significant abnormal degradation of the RCPB since containment is not part of the RCPB: and, therefore, the TS does not satisfy Criterion 1.
0
.g.
This specification is applicable to parameters that are an initial condition
-of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
However, the process variables for which the requirements are applicable (containment design pressure and allowable leakage rates) are not variables that are monitored and controlled during power operation such that process values j
remain within the analysis bounds.
Containment integrity is assured by periodic inspection and testing per TS 3.6.1.1.
Therefore, this specification does not satisfy Criterion 2.
The specification applies to containment leakage rate limits.
Thus, it is applicable to a structure that is part of the primary success path, and which function to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- However, the intent of Criterion 3 1s to capture only those SSC (and supporting systems) that are part of the primary success path of a safety sequence
- analysis, Operability of the containment is assured by a separate LCO 3.6 1.1 and the limits imposed by the leakage rate requirements are neither monitored or controlled during operation nor part of the primary success path of the containment function.
Tnerefore, this TS does not satisfy Criterion 3.
Containment leakage has not been shown to be significant to public health and safety by either operational experience or PSA.
PRAs indicate that offsite release risk 1s dominated by events in which the containment is bypassed, unisolated, or fails structurally.
The technical specification value for overall containment leakage is included in the DCPP Modular Accident Analysis Program (MAAP) model, but contributes only a very small fraction of the total release in the Level 2 Individual Plant Examination (IPE).
Therefore, this TS does not satisfy Criterion 4.
As noted above, the licensee is proposing to add a new surveillance requirement to TS 3.6.1.1.
The SR would s)ecify in addition to the two existing SRs. that containment integrity s1all be demonstrated "by performing containment leakage rate testing except for containment air lock testing, in accordance with the containment leakage rate testing program."
the corresponding requirement in NUREG-1431. Revision 1.
specifies that, to demonstrate containment operable, the licensee shall " perform required visual examinations and leakage rate testing except for containment air lock testing.
in accordance with 10 CFR Part 50. Appendix J. as modified by approved exemptions." The fact that the licensee did not include " required visual examinations" in tN new SR will not relieve the licensee from performing visual examinations, since they are required by 10 CFR Part 50. Appendix J.
With the additions to 3.6.1.1 and the Bases for 3/4 6.1.1. the relocation of TS 3.6.1.2 to a licensee controlled document, namely, the DCPA ECGS is acceptable.
3.8 Containment Structural Integrity in the existing TSs. LC0 3.6.1.6 and SR 4.6.1.6.1 require that the structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate be determined during the shutdown for
is l*
10 -
each Type A containment leakage rate test by a visual inspection of the surfaces.
For the leakage rate test. this BR references SR 4.6.1.2 (discussed in Section 3.7 above), which is being relocated to a license-controlled document.
The containment serves as a barrier to prevent the release of fission products following a LOCA or MSLB inside containment. To mitigate the potential
. consequences of a DBA, it is necessary that the containment structure meet its structural requirements, This specification is intended to detect abnormal degradation of the containment structural elements.
Therefore, this TS requires that the capability of the containment structure to withstand peak accident pressure be demonstrated periodicelly. This TS outlines an appropriate inspection and testing program to demonstrate this capability.
This specification is not ap)11 cable to installed instrumentation that is used to detect, and indicate in tie control room, a significant abnormal-degradation of the RCPB since containment is not part of the RCPB and; therefore, the TS does not satisfy Criterion 1.
This specification is applicable to a design feature (the containment) that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Containment structural integrity is assumed to be available for many DBAs. However, containment structural integrity is not monitored or controlled during plant operation but, rather, via periodic inspections and tests which are performed via TS 3.6.1.1.
Therefore, this specification does not satisfy Criterion 2.
The specification applies to the detection of abnormal degradation of containment structures and therefore to containment structural integrity.
Thus, it is ap)licable to a structure that is part of the primary success path, and whic1 function to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
However, the functional mode addressed by the TS is maintaining the passive, pressure boundary integrity.
The containment system is not required to be a-complete and independent safeguard against a LOCA by itself, but.
functions to contain any fission products released while the ECCS cools the reactor core'.
Therefore, this TS is not required to. ensure the operability of containment and, thus, does not satisfy Criterion 3.
Containment leakage has not been shown to be significant to public health and safety by either operational experience or PSA.
PRAs indicate that risk is dominated by events in which the containment is bypassed, unisolated, or fails structurally.
No DCPP containment vulnerabilities were identified as a results of Supplement 3 to Generic Letter 88-20. The Dt'PP IPE also showed that the overall release frequency per year is dominated by releases due to containment isolation failures. The risk of containment leakage resulting from overpressurization is small because there is a large margin between the DCPP containment design pressure and median containment failure pressure calculated for the IPE. The best estimate was that a small containment failure mode will occur at 2.5 times the design pressure due to cylindrical I-l
t o wall liner plate failure at a penetration insert plate.
Therefore, this TS does not satisfy Criterion 4 The visual inspection requirement does not satisfy any of the four criteria in 10 CFR 50.36.
Except for surveillance of tendons (in plants using prestressed tendons). NUREG-1431. Revision 1. does not include a similar requirement.
The relocation of TS 3.6.1.6 to a licensee-controlled document, specifically the DCPP ECGS is acceptable.
I 3.9 Motor-0perated Valves Thermal Overload Protection and Bypass Devices In the existing TSs. LCO 3.8.4.1 and SR 4.8.4.1 specify the type and frequency of the tests to verify the thermal overload setting for motor-operated valves.
The motor operated valves thermal overload protection provides equipment and distribution system protection from faults or improper operation of other protection devices, in addition to that provided by the design of the distribution system.
The test frequency for trip actuation devices is 18 months ' refueling interval). The test frequency for channel calibration is such that each device is tested at least once per 6 years.
The motor-operated valves thermal overload protection minimizes the potential for an improper setting of a thermal overload. An improper thermal overload setting would prevent a vital piece of equipment from performing its intended function.
This specification does not contain requirements for installed instrumentation that is used to detect. and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy Criterion 1.
The motor-operated valves thermal overload protection helps to preserve the assumptions of the safety analysis by enhancing proper equipment operation.
However, the motor-operated valves thermal overload 3rotection is not a process variable that is an initial condition of a D3A or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not satisfy Criterion 2.
The motor-operated valves thermal overload protection provides equipment and distribution system protection from faults or improper operation of other protection in addition to that provided by the design of the distribution system. However, the motor-operated valves thermal cverload protection is not a structure, system or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes
-the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not satisfy Criterion The thermal overload protection and bypass devices for the MOVs are not explicitly modeled in the DCPP PRA. MOV failure rates are not significantly affected by the presence or absence of the thermal protection and bypass devices, and thus not significant contributors to risk.
Therefore, this TS does not satisfy Criterion 4 1
1 12 The overload protection devices do not satisfy any of the four criteria in 10 CFR 50.36.
Requirements relating to these devices are not included in NUREG-1431. Revision 1.
Relocation of TS 3.8.4.1 to a licensee-controlled document specifically the DCPP ECGS is acceptable.
I 3.10 Containment Penetration Conductor Overcurrent Protection Devices For the containment penetrations with overcurrent protection devices, existing TS 3.8.4.2 and SR 4.8.4.2 specify the types of tests tc be performed and the testing frequency to ensure the devices are operable.
The containment penetration conductor overcurrent protective devices are installed to minimize the possibility that a fault inside containment, or in cabling which aenetrates containment, could damage the electrical penetration and thereby areach the containment structure.
The TS requirements for these devices are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; l
therefore, this TS does not satisfy Criterion 1.
l The containment penetration conductor overcurrent protective devices do help to preserve the assum3tions of the safety analysis by enhancing proper equipment operation. 10 wever, they are not associated with a process variable, design feature or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this requirement does not satisfy Criterion 2.
The containment penetration conductor overcurrent protective devices provide equipment and distribution system protection from faults or improper operation of other protective devices in addition to that provided by the design of the distribution system.
The TS for containment penetration conductor overcurrent protective devices do not apply to a SSC that is part of the primary success path, and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy Criterion 3.
The containment penetration conductor overcurrent protective devices are installed to minimize the potential for a fault in a component inside containment or in cabling which penetrates the containment from damaging the electrical penetration in such a way that the containment structure is breached.
The protective devices have not been shown to be significant to public health and safety by either operational experience or PSA.
The overcurrent 3rotective devices have not been included in the DCPP IPE.
Therefore, t11s TS does not satisfy Criterion 4.
The devices do not satisfy any of the four criteria in 10 CFR 50.36 and are not included in NUREG-1431. Revision 1.
Relocation of TS 3.8.4.2 to the DCPP ECGS. a licensee-controlled document, is acceptable.
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o 3.11 Bases With the removal of the above sections from the TSs, the associated Bases are also being removed or modified.
The Bases being removed include 3/4.3.3.3 on Seismic Instrumentation. 3/4.3.3.7 on Chlorine Detection Systems. 3/4.3.4 on Turbine Overspeed Protection, 3/4.6.1.2 on Containment Leakage, 3/4.6.1.6 on Containment Structural Integrity, and 3/4.8.4 on Electrical Equipment Protective Devices.
As discussed in Section 3.7 above, the limits for allowable containment leakage have been added to the containinent leakage rate testing program. The changes to the Bases are acceptable. The changes discussed herein also necessitate a number of revisions to the index for the TSs.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the California State official was notified of the proposed issuanc,e of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
These amendments change a requiremert with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The NRC staff has I
determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (60 FR 58404). Accordingly, the amendments meet the eligibility ( iteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pur.,aant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
R. Clark Date:
February 3, 1998
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