ML20078S767
| ML20078S767 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 02/16/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078S762 | List: |
| References | |
| NUDOCS 9502270127 | |
| Download: ML20078S767 (6) | |
Text
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UNITED STATES j
NUCLEAR REGULATORY COMMISSION j
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_f WASHINGTON, D.C. Spe00 40M 3
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO' AMENDMENT NO.118 TO FACILITY OPERATING LICENSE NO. NPF-29 E'NTERGY OPERATIONS. INC.. ET AL.
GRAND GULF NUCLEAR STATION. UNIT 1
' i DOCKET N0. 50-416 j
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1.0 INTRODUCTION
By letter dated August 11, 1993, the licensee (Entergy Operations, Inc.),
l submitted a request for changes to the Grand Gulf Nuclear Station, Unit 1 (GGNS) Technical Specifications (TSs). The requested amendment deletes certain accident monitoring instrument limiting conditions for operations (LCOs) from TS Table 3.3.7.5-1 " Accident Monitoring Instrumentation" and deletes the corresponding surveillance requirements (SRs) from~
- Table 4.3.7.5-1, " Accident Monitoring Instrumentation Surveillance l
Requirements." The deleted requirements will be relocated to documents that are controlled by the licensee under the provisions of 10 CFR 50.59. The change is consistent with the format and content of the Improved Standard Technical Specifications (NUREG-1434, Revision 0).
l S>ecifically, the licensee requests the deletion of the specifications from l
tie TSs and their relocation to the Grand Gulf Updated Final Safety Analysis j
Report (UFSAR) and controlled under the provisions of 10 CFR 50.59:
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1.
Relocate the following accident monitoring instrumentation from TS Table 3.3.7.5-1 as well as the LCO requirements to the UFSAR under the licensee's administrative control.
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Drywell/ Containment Differential Pressure Safety Relief Valve Tail Pipe Pressure Switch Indicators Containment Ventilation Exhaust Radiation Monitor Off-gas and Radwaste Bldg. Ventilation Exhaust Radiation Monitor Fuel Handling Area Ventilation Exhaust Radiation Monitor i
Turbine Bldg. Ventilation Exhaust Radiation Monitor Standby Gas Treatment System A & B Exhaust Radiation Monitor 2.
Relocate the corresponding surveillance requirements for these accident monitoring instruments listed in TS Table 4.3.7.5-1 to the
.J UFSAR.
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9502270127 950216
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2.0 BACKGROUND
On February 6, 1987, the Commission issued an interim policy statement on TS improvements, " Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (52 FR 3288). During 1989 through 1992, the utility Owners Groups and the NRC staff developed improved Standard Technical Specifications (STSs) that would establish models of the Commission's policy for each primary reactor type.
In addition, the staff, licensees, and the Owners Groups developed generic administrative and edito-rial guidelines in the form of a " Writers Guide" for TSs, which affords a significant enhancement of human factors considerations and was used throughout the development of licensee-specific improved TSs.
In September 1992, the Commission issued NUREG-1434, which was developed utilizing the guidance and criteria contained in the Commission's interim policy statement.
It was established as a model for developing improved TSs for the BWR/6 plants in general and for the improved Grand Gulf Nuclear Station TSs specifically. NUREG-1434 reflects the results of a detailed review of the application of the interim policy statement criteria to generic system functions, which were published in a " Split Report" issued to the NSSS Owners Groups in May 1988. NUREG-1434 also reflects the results of extensive discussions on various drafts of STSs, so that the application of the TS criteria and the Writers Guide would consistently reflect detailed system configurations and operating characteristics for all NSSS designs. As such, the generic Bases presented in NUREG-1434 provide an abundance of information regarding the extent to which the STSs present requirements which are necessary to protect the public health and safety.
On July 2?,1993, the Commission issued its Final Policy Statement.
- Therein, the Commistion expressed its view that satisfying the guidance in the policy statement also satisfies section 182a of the Atomic Energy Act and 10 CFR 50.36. The Final Policy Statement described the safety benefits of the improved STSs and encouraged licensees to use the improved STSs as the basis for plant specific TS amendments, and for complete conversions to improved STSs.
Further, the Final Policy Statement provided guidance to evaluate the required scope of the technical specifications, and finalized the guidance criteria to be used in determining which of the design conditions and associated surveillances need to be located in the TS. The Commission noted (58 FR at 39136) that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TS, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979). There, the Appeal Board observed:
[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally
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. binding upon the licensee unless and until changed with specific
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Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is 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 i
event giving rise to an immediate threat to the public health and safety.
In accordance with this approach, existing TS requirements which fall within or satisfy any of the criteria in the Final Policy Statement should 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. The Final Policy Statement criteria 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 I
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.
4.
A structure, system, or cor..ponent which operating experience or probabilisticsafetyassepsmenthasshowntobesignificantto public health and safety In its license amend at application, the licensee proposed changes to relocate existing TL vmirements using the Final Policy Statement and r
NUREG-1434 as guidana.
The Commission recently promulgated a proposed change to 10 CFR 50.36, pursuant to which the rule would be amended to codify and incorporate these criteria. This proposed rule clarified the contents of the Bases in NUREG-1434 and specified that only LCOs for Reactor Core Isolation Cooling, Isolation Condenser, Residual Heat Removal, Standby Liquid Control, and Recirculation Pump Trip meet the guidance for inclusion in the TS under Criterion 4.
In the proposed change to 650.36, the Commission specifically requested public comments regarding application of Criterion 4. For the purpose of this evaluation, Criterion 4 has not been applied to add TS restrictions other than those indicated above. See Proposed Rule, " Technical Specifications," 59 FR 48180 (September 20,1994).
Relocated reauirements As summarized above, the Commission's policy statement provides that existing TS requirements which do not satisfy or fall within any of the four specified criteria may be relocated to appropriate licensee-controlled documents.
In the licensee's application, such requirements are generally relocated to the UFSAR and to the TS Bases. The relocated provisions of the existing TS SRs will be relocated to appropriate plant procedures; i.e., operating procedures, maintenance procedures, surveillance and testing procedures, and work control procedures, depending on the reature of the requirements being relocated.
The facility and procedures described in the UFSAR and can only be revised in accordance with the provisions of 10 CFR 50.59, which ensures an auditable and appropriate control over the relocated requirements and any future changes to these provisions. Temporary procedure changes are also controlled by 10 CFR 50.54(a).
As described in more detail in this evaluation, the staff concludes that appropriate controls have been identified for all of the requirements that are being relocated from the licensee's TSs to licensee-controlled documents.
Until incorporated in the UFSAR and procedures, changes to the provisions being relocated from the TSs will be controlled in accordance with the applicable existing procedures that control these documents.
3.0 EVALUATION The licensee has requested the relocation of the LCOs and SRs for the following post accident monitoring (PAM) instrumentation from Table 4.3.7.5-1 to other licensee controlled documents that are controlled under the provisions of 10 CFR 50.59:
Drywell/ Containment Differential Pressure Safety Relief Valve Tail Pipe Pressure Switch Indicators Containment Ventilation Exhaust Radiation Monitor Off-gas and Radwaste Bldg. Ventilation Exhaust Radiation Monitor Fuel Handling Area Ventilation Exhaust Radiation Monitor Turbine Bldg. Ventilation Exhaust Radiation Monitor Standby Gas Treatment System A & B Exhaust Radiation Monitor The primary purpose of the PAM instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis events. The NUREG-1434 instruments that monitor these variables are designated as Type A, Category I, and non-Type A, Category I in accordance with Regulatory Guide 1.97
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-The operability of required accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97.
The PAM instrumentation LC0 requires the operability of Regulatory Guide 1.97, Type A, variables to ensure that the control room operating staff can:
1 Perform the diagnosis specified in the Emergency Operating Procedures j
(EOP). These diagnoses are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs) (e.g., loss of coolant accident (LOCA)); and Take the specified, preplanned, manually controlled actions for which no automatic control is provided, which are required for safety systems to accomplish their safety function.
The PAM instrumentation LCO also requires operability of Category I, non-Type -A, variables. This ensures the control room operating staff can:
Determine whether systems important to safety are performing their intended functions; Determine the potential for a gross breach of the barriers to radioactivity release; Determine whether a gross breach of a barrier has occurred; and Initiate action necessary to protect the public and to obtain an estimate of the magnitude of any impending threat.
Instrumentation that meets the definition of Type A in Regulatory Guide 1.97 satisfies Criterion 3 of the NRC Policy Statement and should be retained in the TSr..
Category I, non-Type A, instrumentation is retained in the TSs because it is intended to assist operators in minimizing the consequences of accidents. Therefore, these Category I, non-Type A, variables are important for reducing public risk.
None of the above instrument requirements, that the licensee proposes to relocate, are categorized'as either Type A, Category I, or non-Type A, Category I.
They do not provide information that is required by the control room operator during and following an accident, nor, do they provide necessary.
support for the operator to take manual action for which no automatic control is provided that is required for safety systems to accomplish their safety function for design basis events. The relocation of instrument functions and SRs will not affect installed control room instrumentation that is used to detect and indicate a significant degradation of the reactor coolant pressure boundary, nor will the relocation affect any structure, system, or component that is required to mitigate the consequences of a design basis accident.
s
. The above relocated requirements relating to installed plant instrumentation i
are not required to be in the TSs under 10 CFR 50.36, and are not required to obviate the possibility of an abnormal situation or event giving rise to an imediate threat to the public health and safety. Further, they do not fall within any of the four criteria set forth in the Comission's Final Policy Statement, discussed in the Introduction above.
In addition, the staff finds i
that sufficient regulatory controls exist under 10 CFR 50.59 to assure y
continued protection of the public health and safety. Accordingly, the staff
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has concluded that these requirements may be relocated from the TSs to the d
licensee's TS Bases or UFSAR, as applicable.
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4.0 STATE CONSULTATION
In accordance with the Comission's regulations, the Mississippi State J
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 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 Comission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public coment on such finding (58 FR 46234). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Comission 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 Comission's regulations, and (3) the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributor:
Paul W. O'Connor Date: February 16, 1995