ML20077H096

Application for Amends to Licenses DPR-42 & DPR-60,revising Tech Spec Sections 3.1.A.2.c.(1) & 3.1.A.2.c.(2) to Improve Reliability of Pressurizer PORVs & Availability of Low Temp Overpressure Protection Sys

ML20077H096
Person / Time
Site:	Prairie Island
Issue date:	06/25/1991
From:	Parker T NORTHERN STATES POWER CO.
To:
Shared Package
ML20077H092	List: ML20077H091 ML20077H096 ML20077H099
References
NUDOCS 9107050163
Download: ML20077H096 (6)

Text

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s er UNITED STATES NUCLEAR RECULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCEET NO. 50 282 50 306 REQUEST 10R AMENDMENT TO OPERATING LICENSES DPR 42 & DPR 60 ,

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LICENSE AMENDMENT REQUEST DATED June 25, 1991 l l

Northern States Power company, a Minnesota corporation, requests authorization i for changes to Appendix A of the Prairie Island Operating License as shown on I the attachments labeled Exhibits A, B, and C. Exhibit A describes the proposed changes, reasons for the changes, and a significant hazards eval- .

uation. Exhibits B and C are copies of the Prairie Island Technical '

Specifications incorporating the proposed changes, This lettar contains no restricted or other defense information, NORTilERN S TE P .R COMPANY By , }p140 h w

' Thomas M Parker Manager Nuclear Support Services

/N)beforemeanotarypublicinandforsaid On this day'of b6 County personally agpeared Thomas M Parker, ManaEer Nuclear Support Services, 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 be-lief the statements made.in it are true.and that it is not interposed for delay.

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k Exhibit A i

Prairie Island Nuclear Generating Plant License Amendment Request Dated June 25, 1991 Evaluation of Proposed Changes to the Technical Specifications Appendix A of r Operating License DPR-42 and DPR CG ,

Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating i Licenses DPR 42 and DPR+60 hereby propose the following changes to Appendix A.

Technical Specifications:

Proposed Changes and Reasons for Change This license amendment request proposes changes to the Prairie Island Technical Specifications provided in response to Generic Letter 90 06.

Generic Letter 90 06 provided NRC Staff gu* dance on Technical Specification changes that should be implemented to improve the reliability of the Pressurizer Power Operated Relief Valves (PORVs) and the availability of the low temperature overpressure protection system.

The Technical Specification changes being implemented in response to Generic Letter 90 06 are described below and the specific wo' ding changes to Technical .

Specification Section 3.1. A.2.c, " Pressurizer Power Operated Relief Valves",

Table TS.4.1 2A, " Minimum Frequencies For Equipment Tests", and the associated Bases are shown in Exhibits B and C.  ;

A- .

Proposed chances'to Technical Specification Section 3.1.A.i.e.(1)

1. The PORV Specification for reactor coolant average temperature greater

. than or equal to 310*F is being revised to delete the limitations on criticality and startup operation. These limitations on criticality and startup operation are mode change limitations. The modified ,

Standard Technical Specifications provided in Attachment A 1 to-Generic Letter 90 06 specified that the provisions of Specification 3.0.4 are  ;

-not applicable to the PORV Specification.

The pressurizer PORV Specification mode change restrictions were incorporated into the. Prairie Island Technical Specifications by License Amendments 91 and 84, approved October 27,.1989 which incorporated a large upgrade into the Prairie Island Technical ,

Specifications. They were added to the PORV Specification as part of

-an effort to standardize the LCOs throughout the Prairie Island Technical Specifications. However, at that time we were not aware of the Standard Technical Specification exception to the requirements of Specification 3.0.4 for pressurizer PORVs. These mode change restrictions were mistakenly incorporated into the Prairie Island PORV_-

Specification as part of' standardized LCO wording. ' '

We are proposing that these pressurizer PORV mode change restrictions be eliminated from the Prairie Island Technical Specifications because the Standard Technical. Specifications state that the provisions 'of _

3.0.4 are not applicable and because the restrictions were mistakenly incorporated by_a previous license amendment.

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2. The requirement for the PORVs to be operable whenever the reactor is critical is being deleted because it is redundant to the requirement  ;

that t.he PORVs be operable whenever reactor coolant average temperature is greater than 310* F. The requirement to be operable above 310*F is

- more restrictive and encompasses the reactor critical condition.

3. The required action with one or both PORVs inoperable is being clarified to differentiate between the actions to be taken if the PORV is inoperable due to excessive seat Icakage or is inoperable for other reasons. These changes include new requirements on maintaining power to the bloch valves and new time limitations / shutdown requirements for the inoperability of PORVs for reasons other than excessive seat leakage. The current action statement does not address the cause of the inoporability or the status of the power supply to the block valves. l

4. The required action with one or more PORV block valves inoperable is being revised to eliminate the option of closing the inoperable block valve and to include new actiona, including new time limitations, .

shutdown requirements and restrictions on automatic PORV operation.

This action would eliminate the option of continued operation with an ,

inoperable PORV block valve. '

B. Proposed chanres to Technical Specification Section 3.1. A.2.c. (2)

1. The requirements for the operability of the Prairie Island low temperature overpressure protection system are currently provided by

• the PORV Specification for reactor coolant system average temperature below 310'F'(Specification 3.1.A.2.c.(2)). - This specification is being revised to provide separate low temperature overpressure protection system specifications for two reactor coolant system low temperature ranges, below 200* F and between 200* F and 310' F.

2. Expanded action statements are being incorporated for the inoperability of PORVs during.either low temperature range. Thesc expanded action-statements' include new time limitations for the inoperability of one or both PORVs and specific requirements for the size of the reactor coolant vent to be utilized'if the PORVs are inoperable longer than the allowed out of service time. The proposed 3 square inch reactor

, coolant system vent opening is based on the 2.956 square inch cross

! sectional. flow area of a pressurizer PORV.

C. ' Proposed changes to Teci t nical Specification Table TS.4.1 IA

1. A' note.is being added to the PORV block valve quarterly surveillance requirement (item 6) which states that a block valve quarterly surveillance need not be performed if the valve has been closed in response to~ proposed action statements 3.1.A.2.c.(1),(b).2 or

.3.1.A.2.c.(1),(b).3.

2. A typographical error is being corrected in the frequency discussion of Item 5. The word "floodes" is being corrected to " flooded".

3. A typographical error is being corrected in the frequency discussion of wr- -- -.-v,- e + -m- + e m v e ,-w , -w----+-e-- >,s- --e - - - - -

Eahlbit A Page 3 of 5 Item 6. The word "Quaterly" is being corrected to " Quarterly".

4. Item 12, which is just a reference to a previously deleted item, is being deleted to provide room to incorporate the new text in item 6.

D. Proposed chnnres to Technical Specification Bases

1. Per the guidance provided in Attachment A-3 to Generic Letter 90 06, information is being incorporated into the PORV Technical Specification Bases to aid in the determination of the operability of the PORVs.

2. Information on the basis of the 3 square inch reactor coolant system vent opening is being incorporated into the Bases for the PORV Technical Specifications.

3. Guidance on the manual operation of the PORVs is being incorporated to support proposed action statements 3.1,A.2.c.(1) (b).4 and 5

4. Editorial changes are being made to the PORV Technical Specification Bases such that consistent terminology is utilized when discussing the low temperature overpressure protection system.

5. Information on the basis for the low temperature overpressure protection system PORV setpoint, and when it is required to be updated, is being incorporated into the Bases for the low temperature overprereure protection Technical Specifications.

6. Per the guidance provided in Attachment B 2 to Generic Letter 90-06, information is being incorporated into the low temperature o/erpressure protection Technical Specification Bases to aid in the determination of the operability of the low temperature overpressure protection system.

7. Several pages of the Bases for Technical Specification Section 3.1 are being renumbered because of the additional text being incorporated by the changes described above.

Safety Evaluation and Determination of Sirnificant llazards Considerations The proposed changes to the Operating License have been evaluated so determine 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.

Under certain conditions, the PORVs are used by operators for recovery from postulated accidents such as a steam generator tube rupture.

Automatic actuation of the PORVs is needed for low temperature overpressure protection of the reactor coolant system. With the exception of the elimination of the reactor criticality and mode change limitations, all of the proposed changes increase the probability the PORV would be available for these functions.

The proposed change eliminating the PORVs to be operable whenever the

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reactor is critical vill not effect the probability or consequenens of an accident. The PORV operability requirements are offectively unenanged by this proposed change, the requirement for the PORVs to be operable above 310* F is inore restrictive and encoropasses the reactor critical requirement.

The proposed change eliininating the snode change limitations from the PORV Specification for reactor coolant average temperature greater than or equal to 310* F could potentially impact the response to a postulated accident. However, the deletion of the mode change limitations is consistent with the guidance provided in Generic Letter 90 06. The limitations on the inoperability of the P0f,Vs provided by the respective action statements, and therefore the PORV availability, would be unaffected by the deletion of the inode change limitations. Additionally, since these inode change restrictions were inistakenly incorporated into the Prairie Island Technical Specifications by a recent' arnendment, they have

- not been utilized in any accident evaluations and their eliinination will not significantly effect the availability of the PORVs to respond to postulated accidents or the consequences of accidents previously l evaluated.

The proposed change of reaintaining power to closed block valves could

- potentially increase the probability of an inadvertent opening of a block valve. The safety innpact is, however, not significant since the proposed changes are.only applicabic if the PORV is inoperable due to excessive seat leakage. 1f the block valve were inadvertently opened only valve leakage would occur, the reactor coolant systern would not undergo a rapid depressurization.

Therefore, based on the conclusions of the above discussion, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously-evaluated.

2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident nreviousiv analyzed.

There are no new-failure modes or mechanisms associated with the proposed

' changes. The proposed changes do not involve any additional testing or any modifica*: ion in the operational limits or physical design of the c involved systems. The proposed license amendment only involves changes to Technical Specification limiting conditions for operation, action-L statements and surveillance requirements which would enhance the availability of the PORVs to perform their design function.

As discussed above. the proposed changes do not result in any significant change in the configuration of the plant, equipment design or equipment use nor do they require any change in the accident analysis methodology..

. Therefore, no different type of accident is created. No safety analyses are-affected. The accident analyses presented in the Updated Safety Analysis Report rereain bounding.

3. The proposed amendment will not involve a significant reduction in the

l. marcin of safety.

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tahlbit A Page 5 of 5 The proposed changes to Technical Specification limiting conditions for operations, action statements and survalliance requirements will enhance the margin of safety provided by the Technical Specifications. The increase in the margin of safety is provided by an increase in the probability that the PORVs would be available if needed for accident  ;

recovery or for low temperature overpressure protection. Therefore, the proposed changes will not result in a significant reduction in the plant's margin of safety.

The pommission has provided guidance concerning the application of the standards in 10 CFR 50.92 for determining whether a significant hazards consideration exists by providing certain examples of amendments that will likely.be found to involve no significant hazards considerations. These ,

examples were published in the Federal Register on March 6, 1986.

The changes to the Praitie-Island Technical Specifications proposed above are equivalent to NRC examples (ii) and (vi), because they either involve changes that constitute additional linitations, restrictions or controla not presently included in the Technical Specifications; or involve changes which may either result in some increase to the probability or consequences of a previously.

analyzed accident or may reduce in some way a safety margin, cet where the

.results of the change are clearly within all acceptance criteria with respect to the' system or component specified in the Standerd Review Plan. Based on this. guidance and the reasons discussed above, we have concluded that the proposed changes do not involve a significant hazards consideration.

Environmental Assessment This license amendment request does not change effluent types or total effluent amounts nor does it involve an increase in power level, Therefore, "this change will not result in any significant environmental impact, I-r h

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