ML20077R515
| ML20077R515 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/10/1995 |
| From: | Wadley M NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20077R511 | List: |
| References | |
| NUDOCS 9501200311 | |
| Download: ML20077R515 (11) | |
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UNITED STATES NUCLEAR REGUIATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISIAND NUCLEAR GENERATING PIANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED January 10, 1995 Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Prairie Island Operating License, Appendix A as shown on the attachments labeled Exhibits A, B, and C.
Exhibit A describes the proposed changes, reasons for the changes, and the supporting safety evaluation /significant hazards determination.
Exhibit B contains current Prairie Island Technical Specification pages marked up to show the proposed changes.
Exhibit C contains the revised Technical Specification pages.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY
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By M. D. Wadley Plant Manage Prairie Island Nuclear Generating Plant On this/d day of_
/ff[eforemeanotarypublicinandforsaid County, personally fppeared Mf D. Wadley, Plant Manager, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.
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LICENSE AMENDMENT REQUEST DATED January 10, 1995 i
Post Accident Monitoring Technical Specifications Conformance to Standard Technical Specifications
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EXHIBIT A Description of the Proposed Changes, The Reasons for Requesting the Changes, and the Supporting Safety Evaluation /Significant Hazards Determination Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to the Facility Operating Licenses and Appendix A, Technical Specifications:
Background
In January 1993 NSP submitted a license amendment request to revise our Core Exit Thermocouple requirements to remedy immediate problems. In that submittal the Core Exit Thermocouple requirements were aligned with the requirements of Standard Technical Specifications. During review of the Core Exit Thermocouple submittal the NRC staff recommended complete conversion of Prairie Island post accident monitoring requirements to Standard Technical Specifications. NSP has compared the existing Prairie Island Post Accident Monitoring Technical Specifications against NUREG-1431, Standard Technical Specifications, Westinghouse Plants, and found many advantages in the Standard Technical Specifications.
Existing Prairie Island Post Accident Monitoring Technical Specifications, Section 3.15, EVENT MONITORING INSTRUMENTATION, are inconsistent in that some important functions are NOT included while some minor functions are included.
The existing Technical Specifications can also be confusing due to the manner in which some requirements are specified. As a result the probability for non-compliance with Technical Specifications is increased. The Action Statements are restrictive when compared to the industry standard embodied in NUREG-1431.
Again this results in applying more resources than the safety importance of the instrumentation would dictate.
For the above reasons conformance to NUREG-1431 would be beneficial.
Incorporation of the proposed changes will reduce the resources applied to low safety value instrumentation and direct more appropriate attention to safety significant post accident monitoring instruments.
Proposed Channes and Reasons for Channes The proposed changes to Prairie Island Operating License Appendix A, Technical Specifications are described below, and the specific wording changes are shown in Exhibits B and C.
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- 1. TABI.E OF CONTENTS: Revise to reflect the deletion of Section 3.15 Subsections A, B and C, revision of the name of the table supporting Section 3.15 and deletion of a table supporting Section 3.15.
Justification: Table of Contents are revised to reflect the changes made in the text and tables of Section 3.15. These are administrative changes made in' support of the technical changes described below.
2.
TECHNICAL SPECIFICATION 3.15 EVENT MONITORING INSTRUMENTATION: Revise applicabilic.y of the Section. Install new general requirements for the Section invoking the table. Delete the existing three subsections of Limiting Conditions for Operation and replace in their entirety with new Action Statements in support of the table. Delete the existing three tables and replace with a single new table.
Justification: The new applicability statement defines the plant MODES to which the Technical Specifications apply. The current applicability statement essentially makes the same statement as the Section Objective and does not define when the instrumentation is required to be operable. When the reactor is in MODES 1 and 2 there is a potential for a design basis event. If Functions can not be monitored as required by the proposed Technical Specifications or by backup means as allowed by the proposed Technical Specifications the remedy is to go to MODE 3 where the reactor is suberitical, the stored energy in the reactor coolant system is lower and the margin to DNB is larger.
The proposed amendments include general requirements which require instrumentation in the Event Monitoring Instrumentation Table to be operable and invoke the Action Statements by reference from the table.
These statements are necessary to provide guidance on how these Technical Specifications are to function.
Standard Technical Specifications allow entering Applicability MODES when LIMITING CONDITIONS FOR OPERATIONS are not met by stating, "LCO 3.0.4 is not applicable." Since current Prairie Island Technical Specifications do not have a requirement comparable to Standard Technical Specification LCO 3.0.4, Specification 3.15.C has been included which explicitly states that entering MODES 1 and 2 is allowable when LIMITING CONDITIONS FOR OPERATION are not met.
The existing three subsections of Limiting Conditions for Operations would be deleted because they are not applicable with the proposed Action Statements to be included in the new Event Monitoring Instrumentation Table. The existing three instrumentation tables would be replaced with a single new table which includes Action Statements.
The proposed new table and 1:s Action Statements, with some exceptions which are discussed below in the Safety Evaluation, are based on the philosophy of NUREG-1431. The Bases for NUREG-1431, Post Accident Monitoring Instrumentation states, " Table 3.3.3-1 in unit specific Technical Specifications shall list all Type A and Category 1 variables identified by the unit specific Regulatory Guide 1.97 Page 2
F analyses, as amended by the NRC's Safety Evaluation Report." Table TF.3.15-1 is the Prairie Island Technical Specification equivalent of NUREG-1431 Table 3.3.3-1. This license amendment request proposes to include instrumentation in Table TS.3.15-1 that is Type A or Category
- 1. This means some instruments will be added to the table and others will be deleted to be consistent with these criteria.
The Action Statements included in NUREG-1431 were developed in cooperative efforts between the NRC, the NSSS vendors and licensees to provide reasonable assurance that plant operators will have the information required to take manual actions for which'no automatic control is provided. The proposed license amendments, with some exceptions, are consistent with the philosophy of NUREG-1431. Specific application and exceptions are discussed further below.
- 3. TECHNICAL SPECIFICATION TABLE TS.4.1-1C. MISCELLANEOUS INSTRUMENTATION SURVEILLANCE REOUIREMENTS: Delete FUNCTIONAL UNITS 22 and 23.
Justification: These FUNCTIONAL UNITS are references to Technical Specification Tables 3.15-2, Event Monitoring Instrumentation -
Radiation, and 3.15-3, Event Monitoring Instrumentation - Reactor Vessel Level. Both of these tables are deleted by the proposed license amendment request. The portions of these tables which are Regulatory Guide 1.97 Type A or Category 1 have been included in the revised Technical Specification Table 3.15-1 for which the surveillance requirements are defined in FUNCTIONAL UNIT 21.
The surveillance frequencies required by FUNCTIONAL UNIT 21 are consistent with the guidance of NUREG-1431, 4.
TECHNICAL SPECIFICATION BASES 3.15 EVENT MONITORING INSTRUMENTATION:
Insert paragraph defining report contents and revise reference for the center core region.
Justification: Many of the action statements require submittal of a report to the Commission when required channels have been inoperable in excess of the allowed outage time. A new paragraph was inserted into the BASES to define the report contents. Technical Specifications define what requirements shall be met and typically do not define how they shall be met. Since this new paragraph is prescriptive on the contents of the report, it was appropriately included in the BASES.
The revised reference for the center core region reflects the changes which are proposed for Technical Specification 3.15.
l Safety Evaluation Back round 6
The primary purpose of post accident monitoring instrumentation is to display plant vs.riables that provide information required by the control room Page 3
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s operators during accident situations. This information provides the support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplith their safety function for design basis accidents.
The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and to assess plant status and behavior following an accident.
Criteria for Selection of Event Monitoring Instrumentation The availability of accident monitoring instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. Instruments which serve these purposes were identified specifically for Prairie Island in the Safety Evaluation of conformance to Regulatory Guide 1.97 as reciuired by Supplement 1 to NUREG-0737.
The instrument channels required to be OPERABLE by this license amendment request include two classes of parameters identified during the Prairie Island implementation of Regulatory Guide 1.97 as Type A or Category 1 variables.
Type A variables are included in the proposed Table of Event Monitoring Instrumentation because they provide the primary information required for the control roo.n operator to take specific manually controlled actions for which no automatic control is provided, and that are required for srfety systems to accomplich their safety functions for design basis accidents.
Category 1 variables are important for inclusion in these proposed Technical Specifications because they provide information:
1.
To determine whether other systems important to safety are performing their intended functions; 2.
To the operators that will enable them to determine the likelihood of a gross breach of the barriers to radioactivity release; and 3.
Regarding the release of radioactive material to allow for early indication of the need to initiate action necessary to protect the public, and to estimate the magnitude of any impending threat.
These key variables identifh 4 specifically for Prairie Island in its Regulatory Guide 1.97 analyses wre submitted to the NRC by letters dated April 15, 1983; September 15, 1983; January 18, 1985; and June 6,1986. In its October 18, 1985 Safety Evaluation Report, the NRC found Prairie Island l
instrumentation acceptable with respect to conformance to Regulatory Guide 1.97. In conformance with this philosophy, instrumentation not on the current Prairie Island Table of Event Monitoring Instrumentation which is either Category 1 or Type A has been incorporated into the Table. Likewise, instrumentation which is neither Category 1 nor Type A has been deleted from the Table.
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s Exceptions to the Guidance of Standard Technical Specifications Standard Technical Specifications were jointly developed by the nuclear industry and the NRC to provide a basis for Technical Specifications contents.
The Bases for the Standard Technical Specifications provide ample justification for their provisions. Those bases generally apply to Prairie Island and will not be repeated in this application. However flexibility is necessary and allowed to accommodate plant specific conditions. This license amendment request proposes exceptions to the specifics of Standard Technical Specifications which NSP believes will meet the intent. These exceptions are addressed as follows.
Applicability Statement Standard Technical Specifications include an Applicability statement that requires post accident monitoring instrumentation during MODES 1, 2, and 3. The Applicability statement in existing Prairie Island Technical Specifications does not explicitly state for which plant operational MODES the post accident monitoring instrumentation is required to be operational. However, the existing action statements require the plant to be placed in hot shutdown (MODE 3) if less than the minimum instrument channels are operable beyond the allowed outage time. Likewise, the requested amendment action statements propose to take the plant to MODE 3 (hot shutdown) when the required channels are it. operable except for those functions with Action Statements 3 as discussed below. Also, the Surveillance Requirements Table currently requires surveillance on these instruments during Modes 1 and 2.
Accordingly the Applicability statement is revised to explicitly state that these instruments are required during MODES 1 and 2.
Taking the plant to MODE 3 when there is less than a full complement of post accident monitoring instrumentation is justified by the reduced probability and consequences of a severe accident in the MODE 3, hot shutdown, condition. At hot shutdown the reactor is suberitical and all of the control rods are inserted. The neutron flux level and power level are decreasing and are orders of magnitude lower than full power operations. In this plant condition, accident severity is reduced due to the reduced stored energy in the reactor core and reactor coolant system along with increased margins to DNB. The core short-life fission products have already significantly decayed away by the time hot shutdown is reached.
Placing the plant in cold shutdown (MODE 4) with less than a full compliment of post accident monitoring instrumentation introduces risks that offset the benefits. The evolutions of taking the plant to cold shutdown involves concomitant challenges to the plant operators and systems. Thus, placing the plant in the hot shutdown mode with these amended Technical Specifications adequately protects the public health and safety if there are inoperable post accident monitoring functions.
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Functions with Action Statement 3 Standard Technical Specifications for Reactor Vessel Water Level and Containment Area Radiation Instrumentation, do not require plant shutdown if both required instrument channel are inoperable (Action Statement 3) because the plant on which the Standard Technical Specifications are based has alternate means of monitoring these functions. This philosophy has been applied to this license application in that functions which have alternate means of monitoring in addition to the required channels invoke Action Statement 3 whsn both required channels are inoperable. More detail on the applicati n of this philosophy is given in the following paragraphs.
Prairie Island has alternate methods for monitoring containment radiation levels as well as the reactor coolant system pressure and containment sump water level. Reactor Coolant System Wide Range Pressure is backed up by two chtanels in addition to the channels which are required by Table TS.3.15-1. Containment Wide Range Sump Water Level is backed up by two channels of Containment Narrow Range Water Level which will provide the control room operators with the information necessary to take post accident actions. Consequently, the requested amendments propose that the plant would not shutdown in the unlikely event that both required channels for these functions were inoperable.
The Regulatory Guide 1.97 evaluation defined containment isolation valves for which position indication is required. These valves also have alternate means of monitoring their position, therefore Action Statement 3 is appropriate in the event the normal position indication is inoperable on both valves in a penetration.
Standard Technical Specifications do not require plant shut down if both channels of Reactor Vessel Level Instrumentation are inoperable due to alternative instrumentation indications which will enable the operators to determine that the reactor has adequate water inventory.
Prairie Island has this capability and therefore the requested amendment does not require shutdown if both channels of Reactor Vessel Level instrumentation are inoperable.
Similarly, this license amendment request proposes to keep the plant operating if both required Hydrogen Monitoring channels become inoperable. The Hydrogen Monitors are backed up by the manual sampling system. Plant procedures require operators to consider taking mitigating actions when the hydrogen concentration in containment reaches 0.5% and requires the containment hydrogen recombiners to be started before the concentration reaches 4.0%. According to the Prairie Island USAR, using the conservative assumptions delineated in Regulatory Guide 1.7, the hydrogen reaches 3.0% at six days following an accident, 3.!* at ten days following an accident and 4.1% at 17 days following an accident. Due to the slow buildup of hydrogen, the operators have sufficient time throughout the course of the event to obtain manual samples of containment to determine appropriate actions.
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Omission of Pressurizer Level Instrumentation from Table TS.3.15-1 The Standard Technical Specification list of post accident monitoring instrumentation includes Pressurizer Level which is consistent with Prairie Island's Regulatory Guide 1.97 evaluation which classified these instruments as Category 1. Prairie Island has three channels of Safety-Related Pressurizer Level Instrumentation which are required by existing Prairie Island Technical Specification 3.5, Table TS.3.5-2A and 4.1, Table TS.4.1-1A. The operability requirements of these existing Technical Specifications are more restrictive than the Event Monitoring Instrumentation requirements.
Including requirements on these components in the Event Monitoring Instrumentation table as well as the Reactor Trip Instrumentation table invites operator error. The operators are used to working with the provisions of Table TS.3.5-2A and could easily forget that the Technical Specifications contain multiple references to these instruments. For these reasons this license amendment request proposes to omit Pressurizer Level Instrumentation from the Event Monitoring Instrumentation table and rely on the more restrictive requirements of Tables TS.3.5-2A and TS.4.1-1A. Pressurizer Level instrumentation will be maintained operable in overy respect to satisfy post accident monitoring requirements under these existing provisions.
Single Loop Temperature Indication Standard Technical Specifications indicate that Reactor Coolant System Hot Leg Temperature and Cold Leg Temperature have two channels per loop. This is contrary to the installed configuration at Prairie Island. The Prairie Island design with a single channel per loop was submitted to the NRC and approved in the Safety Evaluation Leport for compliance with Regulatory Guide 1.97. Alternative indications are available for each application of these instruments which would be defined in the report to the NRC required by the Action Statements if one or two channels were inoperable.
Core Exit Thermocouple Specifications The Core Exit Thermocouples' operability requirements and action statements were recently revised to conform to Standard Technical Specifications and meet the specific design features of Prairie Island. These changes were issued by the NRC on September 7, 1994. The provisions for Core Exit Thermocouples in this proposal are the same as those just issued by the NRC except that the format has been changed along with the rest of the Event Monitoring Table.
Increased Allowed Outage Time j
The proposed Technical Specifications will increase the allowed outage time for the event monitoring instrumentation. With one channel inoperable 30 days are allowed to restore the instrument. This time is based on industry operating experience and takes into account the remaining OPERABLE channel, Page 7
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the passive nature of these instruments, that is, they do not initiate any critical automatic actions, and the low probability of an event requiring post accident monitoring instrumentation during the 30 days.
With two required channels inoperable, most of the functions allow the instruments to be inoperable for seven days before the plant is taken to hot shutdown. Seven days is based on the relatively low probability cf an event requiring post accident monitoring instrument operation and the availability of alternate means to obtain the required information. Requiring restoration of one inoperable channel of the function or going to hot shutdown after seven days limits the risk that the post accident monitoring function will be in a degraded condition should an accident occur.
Changes in Surveillance Requirements Deleting FUNCTIONAL UNIT 22, Post-Accident Monitoring Radiation Instruments, and combining the relevant instruments with FUNCTIONAL UNIT 21 will change the surveillance requirements for the radiation instruments. The proposed technical specifications require Monthly Channel Checks and Calibration each refueling for all event monitoring instrumentation with the exception of the Hydrogen Monitors and Refueling Water Storage Tank Level instrumentation.
Currently the radiation instruments require Daily Channels Checks and Monthly Functional Tests. Deleting radiation instrument Functional Tests and performing Channel Checks Monthly will not reduce the availability of these instruments because their operability is monitored continuously by the plant process computer. These surveillance frequencies conform to the requirements of Standard Technical Specifications for these instruments.
The Hydrogen Monitors' surveillance requirements will not be revised by this amendment and will remain as a separate FUNCTIONAL UU;T with their own surveillance frequencies. These frequencies are based on the manufacturer's recommendations and plant experience and exceed the guidance of NUREG-1431.
Also, the Refueling Water Storage Tank Level instrumentation surveillance requirements will not be revised by this amendment and will remain as a separate FUNCTIONAL UNIT with their own surveillance frequencies.
Conclusion In summary, the proposed Technical Specifications ensure the operability of Regulatory Guide 1.97 Type A and Category 1 variables so that the control room operating staff can take proper, informed accident mitigating actions. Prairie Island specific Regulatory Guide 1.97 Type A and Category 1 variables in the proposed Event Monitoring Instrumentation Table were previously accepted by the NRC. Therefore it is concluded these proposed Technical Specifications amendments do not adversely affect public health and safety, They will, however, improve application of plant resources to safety significant instrumentation.
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Determination of Sirnificant Hazards Considerations The proposed changes to the Operating License have been evaluated to determine i
whether they constitute a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92. This analysis is provided below:
- 1. The proposed amendment will not involve a significant increase in the probability or consecuences of an accident previousiv evaluated The primary purpose of post accident monitoring instrumentation is to l
display plant variables that provide information to the control room i
operators during accident situations. Plant instrumentation was i
evaluated for importance for this function when Regulatory Guide 1.97 i
clasetfications were determined..The Prairie Island Regulatory Guide 1.97 classification of instruments was previously approved by the NRC l
on October 18, 1985. This amendment request proposes to base Prairie Island Technical Specifications on the results of the Regulatory Guide 1.97 evaluation in accordance with the guidance of the industry standard.
Revising the allowed outage time for these instruments will not l
significantly increase the probability or consequences of an accident j
since these instruments do not initiate automatic actions, there are available backup indications and the probability of an event requiring these instruments to be operable is very low.
l Therefore, the probability or consequences of an accident previously evaluated are not affected by any of the proposed amendments.
i 2.
The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously annivzed l
This license amendment request proposes to add instruments to the Technical Specifications which have been previously determined to be l
important for post accident monitoring, and to remove instruments from Technical Specifications which have been previously determined to be less important for post accident monitoring. This amendment ensures the control room operators are provided with the instrumentation required to properly manage an accident situation.
j Therefore, based on the above considerations, the possibility of a new
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or different kind of accident from any accident previously evaluated would not be created.
3.
The proposed amendment will not involve a significant reduction in the margin of safety The post accident monitoring functions do not initiate any automatic l
actions. The instrumentation to be added to the Event Monitoring j
Instrumentation Table was previously recognized through the Regulatory Page 9 i
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Guide 1.97 evaluation process as important for post accident monitoring and would be relied upon if there were an event without this license amendment. Instrumentation to be removed from Technical Specifications was previously recognized to be less important and would not be relied upon very much in an event. Overall, with the trade-off of adding and deleting instrumentation, the margin of safety will not be significantly affected.
The proposed license amendment will increase the allowed outage time for most of the instruments. Again, these instruments do not provide automatic actions, they provide indications for monitoring post accident conditions. All of the instruments have backup or corroborating indications which could be relied upon if the Technical Specifications instruments were inoperable. Also an event requiring use of these instruments has a very low probability. For these reasons the proposed changes in allowed outage time will not result in a significant reduction in the margin of safety.
For these same reasons, the proposed changes in radiation instrument surveillance requirements will not significantly reduce the margin of safety.
Overall, a significant reduction in the margin of safety would not result from this license amendment.
Based on the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.
Environmental Assessment Northern States Power Company has evaluated the proposed changes and determined that:
1.
The changes do not involve a significant hazards consideration, or 2.
The changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or 3.
The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9).
Therefore, pursuant to 10 CFR Part 51 Section 51.22(b), an environmental assessment of the proposed changes is not required.
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