ML20095J792

Application for Amends to Licenses DPR-42 & DPR-60,revising Administrative Controls (Chapter 6) Sections of TS & Other TS Sections Impacted by Chapter 6 Changes to Conform to NUREG-1431,STS Westinghouse Plants,Rev 1

ML20095J792
Person / Time
Site:	Prairie Island
Issue date:	12/14/1995
From:	Wadley M NORTHERN STATES POWER CO.
To:
Shared Package
ML20095J791	List: ML20095J788 ML20095J792 ML20095J795
References
RTR-NUREG-1431 NUDOCS 9512270398
Download: ML20095J792 (23)

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UNITED STATES NUCLEAR REGUIATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PIANT DOCKET NO. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED December 14, 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, C, and D. Exhibit A describes the proposed changes, reasons for the changes, and the supporting safety I evaluation and 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. Exhibit D contains Standard Technical Specification pages marked up to show how the proposed amendments compare.

This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY

.By )YNU  %

M.'D. Wadley Plant Manager Prairie Isla Nuclear Generating Plant On this of m /f/ ore me a notary public in and for said County, personally" appeared M. Di Qadley, 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.

- g-l

HAACIA L LaCORE

NOTARY PUguC4GNNEBOTA i HENNEP91 COUNTY

f MyCumnission EspbesJan.31,2000

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9512270398 951214 PDR ADOCK 05000282 P PDR

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4 5 LICENSE AMENDMENT REQUEST DATED December 14, 1995 4

4 Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Specifications 1

i EXHIBIT A I

t Description of the Proposed Changes, The Reasons for i

Requesting the Changes, and the Supporting Safety i Evaluation /Significant Hazards Determination 4

i j Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders j of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to the Facility Operating Licenses and Appendix A, Technical Specifications:

4 l BACKGROUND 4

} This License Amendment Request revises the Prairie Island Technical Specifications, Administrative Controls Sections i (Chapter 6) and other affected Sections to generally follow the

guidance provided by NUREG-1431, Standard Technical

Specifications, Westinghouse Plants, Revision 1, April 7, 1995.

t While following the Standard Technical Specif3cationa guidance,

the provisions of Generic Letter 93-07, Modification of the i Technical Specification Administrative Control Requirements for i Emergency and Security Plans, have also been incorporated. These

proposed amendments also embody the NRC request, by letter from j Mr. Charles R. Thomas, NRC staff, to Mr. Roger O. Anderson of 4

NSP, dated April 11, 1995, to update the Prairie Island Technical Specifications Monthly operating Reports requirements.

! NUREG-1431, Standard Technical Specifications, Revision 1 i

represents the most recent NRC and industry position on Technical

. Specifications content and appropriately implements the guidance of the 10CFR50.36. Conformance of the Prairie Island Technical

Specifications will enable Prairie Island to simplify the

} Technical Specifications Administrative Controls, make the Technical. Specifications more " operator friendly", and reduce i program and document duplication. Many administrative i requirements in Chapter 6 of the Prairie Island Technical

Specifications are duplicated by existing regulations, the j Updated Safety Analysis Report or the Operational Quality -

Assurance Plan and are needlessly repeated in the Technical ,

Specifications, i

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In that this submittal proposes a "line item improvement" to all of Chapter 6, it is similar to the Rochester Gas and Electric, R.E.Ginna Nuclear Power Plant, application dated May 13, 1994, as supplemented by letters of June 24, and September 27, 1994 and approved by NRC letter dated February 6, 1995.

EEQPOSED CHANGES AND REASONS FOR CHANGES 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. Additionally,

, Exhibit D includes the Administrative Controls Section of the

Standard Technical Specifications marked up to demonstrate how the proposed Prairie Island Section 6 Technical Specifications compare.

1. TABLE OF CONTENTS: Revise to reflect the deletion of TS.3.1.E and its bases, TS.4.4.D, and reformatting of Section 6.

Justification: Table of Contents revised to reflect the

! requested license amendments.

2. Technical Specification 3.1.D.4, MAXIMUM COOLANT ACTIVITY:

Delete reference to annual reporting requirements.

l Justification: The annual report requirement is being deleted 5

from Chapter 6. Further discussion is provided below.

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3. Technical Soecification 3.1.E.3. MAXIMUM REACTOR COOLANT

{ OXYGEN, CHLORIDE AND FLUORIDE CONCENTRATION: Delete this entire section from the Technical Specifications.

I Justification: This section of the Technical Specifications references Specifications 6.2 which is deleted. Since this section would have to be amended to support the proposed i i changes to Chapter 6, this License Amendment Request instead

> proposes, in conformance with Standard Technical I i Specifications, to delete this section in its entirety.

Prudent operation and management of a nuclear power plant l l dictates that the concentration of contaminants in the Reactor Coolant System should be limited to protect the investment in plant equipment. However, the Specification for these limitations do not meet the criteria of 10CFR50. 36 (c) (2) (ii) for inclusion in the Technical )

Specifications and therefore are not a part of the Standard <

Technical Specifications. These chemistry restrictions are important considerations in management of the Prairie Island plant and as such will be retained in licensee controlled implementing procedures and the Updated Safety Analysis Report which provides control under 10CFR50.59. Abnormal Page 2 1

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plant chemistry conditions will continue to be reviewed by plant management in accordance with the current charter of the on site review group.

L 4. Technical Specification TABLE TS.4.1-2B, MINIMUM FREOUENCIES l FOR SAMPLING TESTS: Delete table item 7.

lj l Justification: Item 7 is deleted for consistency with 1 deletion of the maximum Reactor Coolant System oxygen, j chloride and fluoride concentration limitations from j Technical Specifications 3.1.E.3. These samples are required l;

by-a licensee controlled implementing procedure.

I 5. Technical Specification 4.4.D. Residual Heat Removal SysteBi

! Delete this section in its entirety.

j Justification: The provisions of this Specification are encompassed by new Specification 6.5.B.

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! 6. Technical Soecification 4.6, PERIODIC TESTING OF EMERGENCY j POWER SYSTEM: Delete Specification 4.6.A.1.c for sampling of i diesel fuel, i

! Justification: The requirements of this specification have j been relocated to a new program requirement in Section 6.5.

l 7. Technical Specification 5.1, DESIGN FEATURES: Delete the l 3

description of emergency procedures for floods and j earthquakes and the criteria for plant shutdown in response to these events.

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l Justification: Proposed Section 6.4 will reference j Attachment A to Regulatory Guide 1.33 which identifies, j " Procedures for Combating Emergencies and Other

! Significant Events". One of the items, Acts of Nature, j includes floods and earthquakes. Thus, through this

mechanism, the Technical Specifications still require the licensee to maintain these procedures. Many of the i procedures listed in Regulatory Guide 1.33 require plant j shutdown without a concomitant requirement in the 1 l

Technical Specifications. Likewise, plant shutdown

criteria are not required or appropriate in the Technical l Specifications for floods and earthquakes.

The original Safety Evaluation Report for Prairie Island issued September 28, 1972, as Supplemented March 21, 1973, j April 30, 1973 and May 31, 1973 specifically stated that the j

Prairie Island Technical Specifications will contain a J i provision to shut down the plant at a flood elevation of +692

feet MSL at the plant site. As stated above, this

{ Specification is not necessary and NSP proposes to remove it through this License Amendment Request.

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Furthermore, these requirements do not meet the guidance of 10CFR50.36 for inclusion in the Technical Specifications and l thus are appropriately deleted. l

8. Technical Specification ADMINISTRATIVE CONTROLS, Section 6.1, Resoonsibility: Revise existing wording and add new paragraphs to conform to Standard Technical Specifications.

Justification: The first paragraph of this section is revised to conform to Standard Technical Specifications except that the title " plant manager" is used here and throughout these proposed amendments in lieu of " Plant Superintendent" since it is the title currently in use at Prairie Island for the position with overall responsibility. The statament that the PM is responsible for the Fire Protection Program was deleted since the paragraph already states that, "The plant manager shall be responsible for overall unit operation . . . " which includes the Fire Protection Program.

The capitalization on the title " plant manager" has been changed to lower case to indicate that this is a generic title associated with the responsibilities rather than plant specific title. Throughout these proposed amendments, generic titles describing responsibilities are proposed in lieu of plant specific titles. Specifying management

' positions by title within the Technical Specifications creates an NRC and licensee burden when titles are changed or responsibilities shifted. Proposed Technical Specification 6.2.A.1 states, " . . . the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be .

documented in the Updated Safety Analysis Report." The intent of this statement is to allow the licensee to make title changes and provide appropriate documentation in the Updated Safety Analysis Report in accordance with the requirements of

10CFR50.59. Summaries of Updated Safety Analysis Report

changes are provided to the NRC on a periodic basis which

affords further NRC review.

i l The second paragraph was added in conformance with the

Standard Technical Specifications and provides further

description of the plant manager's responsibilities.

The third paragraph, added in conformance with the Standard Technical Specifications, provides requirements for the ,

3 control room command function.

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9. Proposed Technical Soecification ADMINISTRATIVE CONTROLS.

Section 6.2, Oraanization: Revise to conform to the Standard Technical Specifications as discussed below.

Justification:

A. Onsite and Offsite Oraanizations This section conforms to the Standard Technical Specifications except as follows:

When both units are collectively under consideration, " plant" has been substituted for " unit."

a 6.2.A.1 The phrase, " . . ., including the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, .

. . " has been added. As discussed above, this phrase will allow the licensee to change organizational titles without

. burdening the NRC and NSP with Technical Specifications changes. Title changes will be documented in the Updated Safety Analysis Report and evaluated in accordance with 10CFR50.59. These evaluations are available for onsite NRC inspection and summaries are submitted to the NRC in accordance with 10CFR50.71(e).

6.2.A.3 A title for the position characterized in the Standard Technical Specifications as "a specified corporate

executive position" has been defined as "A corporate vice president" in lower case letters. This establishes the level of the position within the corporation with overall resposibility and allows title changes without concomitant i License Amendment Requests as discussed previously.

B. Plant Staff The term " plant staff" in lieu of " unit staff" is used in this section since the Prairie Island plant comprises two units which are governed by a common Technical Specifications. Discussion of changes, comparison with

Standard Technical Specifications and justifications are as follows

6.2.B.1 In some instances Prairie Island may assign licensed operators to perform outplant operations. Rather than specify assignment of non-licensed, this amendment proposes use of the term " operator to perform non-licensed duties". Since the current Prairie Island Technical Specifications shift staffing commitments require a minimum of two outplant operators when both units are shutdown or defueled, the Standard Technical Specifications guidance for two unit plants has not been incorporated.

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Current Technical Specifications require one additional outplant operator when one or both units are operating, thus this paragraph has been revised to preserve the current level of staffing. Current Technical Specifications also require a minimum of two senior reactor operators and two licensed reactor operators if one unit is operating and one unit is shutdown, therefore, an additional sentence is provided to preserve this licensed operator staffing level.

Current Technical Specifications requirements for shift staffing as specified in Table TS.6.1-1 have been deleted since these requirements are mandated by 10CFR50.54 (m) (2) (i) or specified in the text of the Technical Specifications.

Table Technical Specifications.6.1-1 has been deleted.

. 6.2.B.2 This paragraph specifies a licensed operator shall be in the control room rather than a Reactor Operator since a Senior Reactor Operator could also fulfill this requirement.

Since Prairie Island is a two unit plant, "the unit" has been replaced with "either unit" to assure proper staffing is maintained. This new specification replaces the current Technical Specifications which follows paragraph 6.2.B.2 in J

Exhibit B.

6.2.B.3 This specification conforms to the Standard Technical Specifications. The current Technical Specifications, following new paragraph 6.2.B.3, has been deleted since it is mandated by 10CFR50.54 (m) (1) and (2) (1) .

6.2.B.4 In accordance with the provisions of the current Technical Specifications, the term "an individual qualified in radiation protection procedures" is used in lieu of

" health physics technician." "a reactor" is used instead of "the reactor" since Prairie Island is a two unit site. In accordance with Standard Technical Specifications guidance, provision for temporary vacancy of this position has been included.

6.2.B.5 This statement will take the place of current Technical Specifications 6.1.E which will be deleted. Prairie Island currently uses a staff overtime control program which slightly differs from the NRC Policy Statement on Working Hours (Generic Letter 82-12). This program was previously reviewed and approved by NRC Safety Evaluation under cover letter dated March 17, 1983 from Domenic Vassallo to D.M.

Musolf. The proposed wording is consistent with one of the options presented in the Standard Technical Specifications.

Current Technical Specifications 6.2.B.5 has been deleted since it is redundant to requirements in 10CFR50.54(m)

(2)(iv).

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6.2.B.6 A generic title has been inserted and "SRO" license is used rather than " current license" for consistency with Standard Technical Specifications. Current Technical Specifications 6.1.C.7 conditions have been deleted since, "or assistant operations manager", was included in this paragraph in accordance with the guidance of Standard Technical Specifications.

6.2.B.7 Qualifications for STA personnel are not included here since they are addressed in Specification 6.3, Plant Staff Qualification 2. For consistency with current Technical Specifications, a statement has been added to define when personnel performing STA functions are required.

Current Technical Specifications 6.1.D was relocated to form the substance for new Technical Specifications 6.3. Technical Specifications 6.1.E is deleted as discussed above in 6.2.B.S.

10. Current Technical Specification ADMINISTRATIVE CONTROLS, Section 6.2, Review and Audit: Delete entire current section.

Justification: This section contains requirements for the l Safety Audit Committee (offsite review function including i independent auditing), the Operations Committee (onsite review function) and Maintenance Procedures. In j conformance with the guidance of the Standard Technical i Specifications, this section is proposed to be deleted in

! its entirety. NSP has committed to comply with the i requirements of ANSI N18.7-1976, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, as defined in the Operational Quality

Assurance Plan.

i' Prairie Island will continue to maintain offsite and onsite review functions, independent auditing and procedure control in accordance with the requirements of ANSI N18.7-1976, the

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Operational Quality Assurance Plan, Updated Safety Analysis

Report and licensee controlled procedures. NRC control over these requirements will continue to be maintained through the

provisions of 10CFR50.54 (a) (3) and 10CFR50.59.

11. Current Technical Specification ADMINISTRATIVE CONTROLS, Section 6.3 Soecial Insoections opd Audits: Delete this section in its entirety.

Justification: The special audits required by this Specification are contained in the Prairie Island Fire Protection Program and are unnecessary in the Technical Specifications. The contents of this section do not meet the guidance for inclusion in the administrative controls

Technical Specifications contained in NRC letter dated Page 7 i

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October 25, 1993, " Content of Standard Technical Specifications, Section 5.0, Administrative Controls".

Specifically these audits are contained in a licensee program and are not required in the Technical Specifications to assure safe operation of the plant.

12. Proposed Technical Specification ADMINISTRATIVE CONTROLS.

Section 6.3 Plant Staff Oualifications: The requirements of current Technical Specifications 6.1.D shall be relocated to Technical Specifications 6.3 and revised as discussed below.

Justification: These requirements from current Technical Specification 6.1.D generally conform to Standard Technical Specifications except the following existing requirements have been retained. Prairie Island, through the existing Technical Specifications and the operational Quality Assurance Plan, has committed to Regulatory Guide 1.8, Revision 1, September 1975 and ANSI N18.1-1971 which is also endorsed by the Regulatory Guide.

NRC license requirements for the operations manager in new Technical Specification 6.2.B.6 are referenced for clarity.

The current Technical Specifications qualifications for personnel who perform the function of STA have been retained.

These were previously defined for the position of Shift Manager, however, in keeping with the philosophy of using generic titles as discussed above, the specific title has been removed from the Technical Specifications.

13. Current Technical Specification ADMINISTRATIVE CONTROLS.

Section 6.4 Safety Limit Violation: Delete this section in its entirety.

Justification: Prairie Island requested the requirements of this section be relocated to Technical Specification 2.2 in License Amendment Request dated May 4, 1995, 4 " Pressurizer Safety Valves and Main Steam Safety Valves Lift Setting Tolerance Change and Safety Limit Curve Changes." Further justification for this change can be

found in that submittal.

14. Proposed Technical Specification ADMINISTRATIVE CONTROLS.

Section 6.4 Procedures and current Technical SDecification Section 6.5: Revise to conform to Standard Technical ,

Specifications as discussed below.  !

Justification: Proposed Technical Specifications 6.4.A

, through 6'.4.E and current Specifications 6.5.A through l 6.5.D are addressed in this section.  !

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l t The current Specification 6.5 lead-in paragraph has been

! revised to conform to the Standard Technical Specifications 6.4 lead-in paragraph. In conformance with Standard Technical specifications, current requirements for Operation committee

, review and plant management approval of procedures have.been deleted. Procedure review and approval will. continue to meet the requirements of ANSI N18.7 as specified.in the j operational Quality Assurance Plan and licensee' controlled

procedures.

l The essence of this proposed section is che commitment in new

Specification 6.4.A to institute the procedures recommended

! by Appendix A to Regulatory Guide 1.33. (This is not a i commitment to the main body of Regulatory Guide 1.33 but only to the applicable procedures recommended in Appendix A to the

, Regulatory Guide.) This appendix to the Regulatory Guide i envelopes most of the other procedures presently required by

current Technical Specification 6.5.A including procedures j for

plant startup, operation and shutdown procedures; fuel handling; malfunctions and abnormal conditions; surveillance and testing; plant emergencies; and natural phenomenon such 3 as earthquakes and floods. On the basis of this commitment, the following current Technical Specifications are deleted:

6.5.A items 1 through 4 and 6, 6.5.B lead-in paragraph; and j 6.5.C.

! Current Technical Specifications item 6.5.A.5 has been i deleted since the requirement for written procedures which implement the Emergency Plan are specified in 10CFR50.54(q)

! and 50.54(t), and 10CFR50, Appendix E, Section V and

therefore further specification in the Technical j Specifications is unnecessary. Deletion of this requirement j is consistent with the Standard Technical Specifications.

New Technical Specifications 6.4.B requires emergency j~ operating procedures to implement NUREG-0737, Supplement 1

, which is the existing Prairie Island commitment as specified 1 in NSP's Operational Quality Assurance Plan.

Current Technical Specification 6.5.A.8 was incorporated into j new Technical Specification 6.4.C and is consistent with

Standard Technical Specifications except that " quality ,

control" is used as currently specified in the Prairie Island '

Technical Specifications in lieu of " quality assurance" as i

stated in Standard Technical Specifications. l 1

The provisions of current Technical Specification 6.5.B lead-  !

l in paragraph, in addition to being required by Appendix A to l i Regulatory Guide 1.33, are also required by 10CFR20.1101 and therefore do not need to be stated in the Technical  !

l Specifications.

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i Current Technical Specification 6.5.B.1, High Radiation Areas, is deleted in favor of revised wording proposed in a License Amendment Request dated July 17, 1995 which is under review at the NRC at the time of this submittal. Any changes to the proposed Technical Specifications arising from the NRC review will be incorporated into these proposed Technical Specifications amendments. These requirements have been relocated to a new section 6.7.

Current Technical Specifications 6.5.B.3 has been deleted

since the silver zeolite program for measuring airborne iodine in the plant is embodied in the Prairie Island Emergency Plan. Changes to the Emergency Plan are controlled

by 10CFR50.54(q) and 50.4, therefore, this requirement in the Technical Specifications is unnecessary. Deletion of this j specification is consistent with Standard Technical 1 Specifications.

Current Technical Specification 6.5.C, as discussed above, is deleted since these requirements are covered by the commitment to provide procedures recommended by Appendix A to Regulatory Guide 1.33.

Deletion of current Technical Specification section 6.5.D was addressed in the Radiological Effluents Technical Specifications License Amendment Request dated July 17, 1995 which is under review at the NRC at the time of this submittal. Any changes to the proposed deletion arising from i the NRC review will be incorporated into these proposed amendments.

15. Proposed Technical Soecification ADMINISTRATIVE CONTROLS.

Section 6.5 Procrams and Manuals: Revise to conform to standard Technical Specifications as discussed below.

Justification: Proposed Technical Specifications 6.5.A through 6.5.L and current Specifications 6.5.E through G 1 are addressed in this section.

Current Technical Specification 6.5.E, Offsite Dose Calculation Manual, has been relocated to proposed Technical Specification 6.5.A. Revised wording was submitted to the NRC

in the Radiological Effluents Technical Specifications License Amendment Request dated July 17, 1995 which is under review at the NRC at the time of this submittal. Any changes l to the proposed Technical Specifications arising from the NRC 1 i review will be incorporated into these proposed amendments.

3 Current Technical Specification 6.5.B.2 has been relocated to l proposed Technical Specification 6.5.B, " Primary Coolant

Sources outside Containment", with minor wording changes proposed for consistency with Standard Technical Page 10

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. Specifications and some specific systems included in the i program identified. Reference to the NSP letters which form l

the basis for this program ~has been deleted since these i letters are a matter of record. The program will continue to meet the requirements stipulated in the proposed Technical Specifications.

j Current Technical Specification 6.5.B.4 has been relocated to proposed Technical Specification 6.5.C, entitled, " Post i Accident Sampling", and minor wording changes are proposed

for consistency with Standard Technical Specifications.

! Proposed Technical Specification 6.5.D, Radioactive Effluent

Control Program, will comprise the Radioactive Effluent l Controls Program previously proposed as section 6.5.H in the Radiological Effluents Technical Specifications License i Amendment Request dated July 17, 1995 which is unaer review 4 at the NRC at the time of this submittal. Any changes to the I

proposed Technical Specifications arising from the NRC review will be incorporated into these proposed amendments.

In conformance with Standard Technical Specifications, proposed Technical Specification 6.5.E, Component Cyclic or ,

j Transient Limit, will replace the requirements of current i Technical Specification 6.6.B.8 which will be deleted. This )

proposed Technical Specification is intended to maintain the 1

current Prairie Island level of commitment and does not  !

decrease inspection or testing requirements. )

. Section 6.5 of the proposed Technical Specifications does not ,

include a program for Pre-Stressed Concrete Containment I l Tendon Surveillance Program. A Pre-Stressed Concrete Containment Tendon Surveillance Program is not appropriate j for the Prairie Island since it does not have prestressed tendon containment design.

j This License Amendment Request does not propose a Reactor

Coolant Pump Flywheel Inspection Program in Chapter 6 since i the provisions of current Technical Specification 4.2 are l being retained at this time. When complete conversion of the

Prairie Island Technical Specifications to Standard Technical

" Specifications are made, this program will be incorporated at

that time. Section 6.5.F will be reserved for the Reactor

} Coolant Pump Flywheel Inspection Program so the order of J programs can be maintained consistent with Standard Technical Specifications with minimal impact.

I This License Amendment Request does not propose an Inservice j Testing Program in Chapter 6 since the provisions of current Technical Specifications 4.2 are being retained at this time. l' 4

When complete conversion of the Prairie Island Technical i Page 11 4

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Specifications to Standard Technical Specifications are made, j this program will be incorporated at that. time.- Section 6.5.G

will be reserved for the Inservice Testing Program so the order of programs can be maintained consistent with Standard-j Technical Specifications with minimal impact.

i This License Amendment Request does not propose a Steam Generator Tube Surveillance Program as indicated in the Standard-Technical-Specifications and instead proposes,to-

continue'to follow the requirements of current Technical Specification 4.12. Currently the industry is working on

! generic program requirements. An attempt to include such a program here would be premature. Section 6.5.H will be reserved for the Steam Generator Tube Surveillance Program if

! one-is proposed for inclusion in the Technical Specifications

in the future. In this way the order of programs can be maintained consistent with Standard Technical Specifications with minimal impact.

i This License Amendment Request does not propose a Secondary Water Chemistry Program as indicated in the Standard 3

Technical Specifications. Prairie Island has aggressively 1

administered a secondary water chemistry program since the startup of the units without any Technical Specification requirements. Huge safety and economic incentives dictate the 1 prudence of maintaining the steam generators through all available means including secondary water chemistry. Steam l

generator degradation has resulted in power reductions, I steam generator replacement and/or shutdown of other nuclear

. plants. Due to the Prairie Island steam generator maintenance program, including secondary water chemistry, the Prairie Island steam generators continue to provide safe performance and are projected to continue to do so for many years.

Addition of Secondary Water Chemistry Program Technical i Specifications would not further improve plant performance or provide any benefit for the health and safety of the public.

This Technical Specification does not meet the NRC criteria 4

for administrative controls defined in the NRC October 25,

, 1993 letter to the Owner Groups Technical Specifications Committee Chairmen. Past performance at Prairie Island demonstrates that this Technical Specification is not i required to assure operation of the plant in a safe manner.

, Technical Specification 6.5.I is reserved for the Ventilation l Filter Testing Program when Prairie Island converts to the Standard Technical Specifications. Such a program is not proposed in this License Amendment Request since it would require extensive revision of current Technical Specifications 4.4, 4.14 and 4.15 and significantly expand

and complicate the scope of this License Amendment Request.

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i Proposed Technical Specification 6.5.J, Explosive Gas and Storage Tank Radioactivity Monitoring Program, will comprise the Explosive Gas and Storage Tank Radioactivity Monitoring Program previously prcposed as section 6.5.I in the Radiological Effluente Technical Specifications License Amendment Request dated July 17, 1995 which is under review at the NRC at the time of this submittal. Any changes to the proposed Technical Specifications arising from the NRC review will be incorporated into these proposed amendments.

In accordance with the guidance of Staindard Technical Specifications, proposed Technical Specification 6.5.K, Diesel Fuel Oil Testing Program, will replace the requirements of Current Technical Specifications 4.6.A.1.c which will be deleted. This proposed Technical Specification differs from Standard Technical Specifications in that the Current Technical Specifications reference to ASTM D975-77 and specific listing of tests is retained. This Specification is intended to mainthin the current Prairie Island level of commitment and does not in any way decrease inspection or testing requirements.

Section 6.5.L, Technical Specification Bases Control Program, is a new program which will allow changes to the Technical Specifications BASES without prior NRC review and approval under the specified conditions. This program generally follows the guidance of Standard Technical Specifications.

Section 6.5 does not, at this time, propose a Safety Function Determination Program as provided in the Standard Technical Specifications. A Fafety Function Determination Program would likely be very beneficial, however, implementation of such a program would require extensive revision of the LCO portion of the Technical Specifications. Such a revision is beyond the intended scope of this License Amendment Request and therefore is not currently proposed. This Program will be included with a complete conversion of the Prairie Island Technical Specifications to the Standard Technical Specifications.

Current Technical Specification 6.5.F, Security, has been deleted since the requirement for written procedures which implement the Security Plan are specified in 10CFR50.54 (p) (1) and 10CFR73.55(b) (3) and therefore unnecessary in the Technical Specifications. The Security Plan includes provision for review and approval of changes, thus, deletion of these requirements implement the recommendations of Generic Letter 93-07. Deletion of Security Plan procedure requirements also is consistent with the Standard Technical Specifications.

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I Current Technical Specification 6.5.G, Temporary Changes to Procedures, has been deleted. Provision for temporary changes to procedures will be included in the Operational Quality Assurance Plan.which will continue to be controlled by the NRC through 10CFR50. 54 (a) (3) .

. 16. Current Technical Soecification ADMINISTRATIVE CONTROLS.

Section 6.6 Plant Operatina Records: Delete this section in

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its entirety.

l Justification: Prairie Island maintains records in j accordance with the requirements of ANSI N45.2.9 in j accordance with the Operational Quality Assurance Plan 2

commitment and the record retention provisions included in each individual procedure, program and manual. Also retention of some records is addressed in various

! regulations such as 10CFR20, Subpart L and 10CFR50.71.

Thus, current Technical Specification 6.6 unnecessarily duplicates other existing controls.

Furthermore, specification of records for retention does not meet the criteria for administrative controls Technical

. Specifications delineated in the NRC letter to Technical i specification Committee Chairmen, dated October 25, 1993.

{ Specifically, listing of required records are not necessary j to assure operation of the facility in a safe manner.

Retention of these records provides documentation J retrievability for review of compliance with requirements and

assure operation of the facility in a safe manner, as

- activities described in these documents have already been

- performed. Removal of this Section will simplify the 3

Technical Specifications and make it more convenient for the

use of the plant operators. Therefore, following the guidance i of the Standard Technical Specifications, this section is

! deleted in its entirety.

! 17. Proposed Technical Specification ADMINISTRATIVE CONTROLS.

Section 6.6 Reoortina Reauirements. Current Technical

[ Soecification ADMINISTRATIVE CONTROLS. Section 6.7 ReDortina j Reauirements: Revise as shown in Exhibits B and C.

1 l Justification: Current Technical Specification 6.7.A.1.a, Occupation Exposure Report and the associated footnote, is revised to conform to Standard Technical Specifications.

l Current Technical Specification 6.7.A.1.b, Report of Safety t 4

and Relief Valve Failures and Challenges, is deleted in conformance with the Standard Technical Specifications which include safety and relief valve failures in the monthly l report (further discussion below). ,

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2 Current Technical Specification 6.7.A.1.c, Primary Coolant

Iodine Spike Report, is deleted since this report is no longer necessary. When Prairie Island started operation, the commercial nuclear industry did not have good data on core

! iodine levels, so, this report provided baseline data to the NRC for use in postulating accident iodine releases. With the-i maturing of the industry this data is well established and

. this report is unnecessary.

! Startup Reports as required by Technical Specification

6.7.A.2 are required when an operating license is received,

i. plant power level is increased, fuel of a different design or manufacturer is installed or modifications are performed j which significantly alter the nuclear, thermal, or hydraulic l performance of the plant. All of these plant changes are i accompanied by specific NRC authorization and requirement for

- a report would appropriately be. addressed in the concomitant

.; Safety Evaluation Report. Accordingly this report requirement j is deleted from the Prairie Island Technical Specifications.

i Current Technical Specification 6.7.C.1, Annual Radiation Environmental Monitoring Report, has been relocated to l proposed Technical Specification 6.6.B, with the same title.

3 The Radiological Effluents Technical Specifications License Amendment Request dated July 17, 1995 which is under review at the NRC at the time of this submittal proposed revised wording for Current Technical Specifications 6.7~.C.1 and proposed deletion of current Technical Specification 6.7.C.2, Environmental Special Reports. Any changes to these proposed 4

Technical Specifications arising from the NRC review will be incorporated into these proposed amendments. All existing i text in current Technical Specification 6.7.C.1 and 2 is deleted.

t

! Current Technical Specification 6.7.A.4, Annual Radioactive j Effluent Report, has been relocated to proposed Technical

Specification 6.6.C, Radioactive Effluent Report. Revised

wording was submitted to the NRC in the Radiological l Effluents Technical Specifications License Amendment Request dated July 17, 1995 which is under review at the NRC at the i time of this submittal. Any changes to the proposed Technical

. Specifications arising from the NRC review will be incorporated into these proposed amendments. All existing text in current Technical Specification 6.7.A.4 is deleted.

! Monthly Operating Report requirements, proposed Specification 1 6.6.D, have been conformed to Standard Technical l Specifications except that the requirement to include documentation of challenges to the pressurizer power operated

, relief valves or pressurizer safety valves has not been i

included. Transients which challenge these valves will likely l be significant enough to be reviewed by the NRC Resident l Page 15 1

. ~ - . - . . . . . - . . .

l Inspector and would likely require reporting under the ,

provisions of 10CFR50.73. Activity of these valves has ,

received heightened industry attention for the last 16 years.

Plant operators and support personnel receive extensive )

training on plant transients including conditions which will operate the pressurizer power operated and safety valves. .

Reporting of valve activity on a monthly or annual basis is unnecessary. Revision of this specification also incorporates i the NRC request by letter from Mr. Charles R. Thomas to Mr. l Roger O. Anderson dated April 11,.1995 to delete the l incorrect mailing address for the report.  :

Current Technical Specification 6.7.A.5, Annual Summaries of Meteorological Data, is deleted since it does not meet the criteria for administrative controls specifications as delineated in the NRC letter from Mr. William T. Russell to owner Group Technical Specifications Committee Chairmen, dated October 25, 1993. These reports are not required for safe operation of the plant and a report has never been requested by the NRC as provided in the current Technical Specifications. Baseline meteorological data is included the Updated Safety Analysis Report and current data is maintained l as plant lifetime records if it should ever be needed.

Removal of this requirement is consistent with the Standard  !

Technical Specifications.

The Core Operating Limits Report (COLR) requirements have been revised to conform to the Standard Technical Specifications and proposed as Technical Specification 6.6.E.

This License Amendment Request does not propose a Reactor Coolant System PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) as indicated in the Standard Technical Specifications and instead proposes to continue to follow the requirements of current Technical Specifications 3.1.B. Preparation of the analyses and reports required for this report would require extensive effort at this time. Prairie Island is approaching the end of 20 EFPY for which the current curves are applicable and new analyses will be required. When the current curves are evaluated for continuing plant operation, a PTLR can be generated and the appropriate changes to the Technical Specifications requested.

An EDG Failure Report is not proposed. The requirement for a special report following four or more valid failures of an individual emergency diesel generator in the last 25 demands was not added since the requirement is not specified in the current Prairie Island Technical Specifications. Any required report can be adequately controlled by the licensee administrative controls.

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. - . . - - . ~ . .-- . - ~ . - . - - . - .

Provision for reporting of Post Accident Monitoring instrumentation failures is not proposed here because those reporting requirements are included in current Technical Specification Section 3.15 which was issued November 9, 1995 in License Amendment 121.

i Steam Generator Tube Inspection Report requirements are not

proposed at this time in Section 6.6 since the current Specification 4.12, Steam Generator Tube Surveillance, E.

Reports, requirements have been retained.

Current Technical Specification 6.7.B, Reportable Events, is deleted. Reportable Events are required to be reported to the NRC in accordance with 10CFR50.73. Inclusion in the Technical 4

Specifications is redundant and unnecessary. Removal from Technical Specifications is consistent with the Standard j Technical Specifications.

Requirements for Other Environmental Reports in Specification 4 6.7.C.3 do not meet the requirements for administrative i controls Technical Specifications as delineated in the NRC letter from Mr. William T. Russell to Owner Group Technical Specifications Committee Chairmen, dated October 25, 1993.

i These reports are not required for safe operation of the

! plant. Removal is consistent with the Standard Technical Specifications.

Distribution of reports as required by Technical l Specifications 6.7.D, Special Reports, is a matter of law and ,

I working relationships with the NRC and is unnecessary in the l Technical Specifications. Deletion of these requirements is consistent with Standard Technical Specifications.

! 18. Proposed Technical Soecification ADMINISTRATIVE CONTROLS.

l Section 6.7 Hiah Radiation Areas: Revise as proposed in i License Amendment Request, Radiological Effluent Technical Specifications Conformance to Standard Technical l Specifications and Generic Letter 89-01, dated July 17, 1995.

l Justificationi Proposed Section 6.7 will comprise the high t radiation area specifications previously proposed as

section 6.5.B.1 in the Radiological Effluents Technical Specifications License Amendment Request dated July 17,

1995 which is under review at the NRC at the time of this '

submittal. Any changes to the proposed Technical

! Specifications arising from the NRC review will be

incorporated into these proposed amendments.

i 19. Technical Specification Bases. 3.1.E. REACTOR COOLANT SYSTEM.

Maximum Coolant Oxvaen. Chloride, and Fluoride Concentration:

i Delete this basis in its entirety.

I

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Justification: Since LCO 3.1.E is proposed to be deleted this basis no longer supports any specification and therefore is unnecessary.

20. Technical Specification Bases, 4.4 CONTAINMENT SYSTEN TESTS:

Delete the paragraphs relating to Residual Heat Removal System tests.

Justification: The provisions of Specification 4.4.D were superseded by specification 6.5.B and therefore were deleted.

The paragraphs in Section 4.4 supporting Specification 4.4.D are deleted since they no longer provide the basis for a specification.

SAFETY EVALUATION The NRC defined the criteria for-administrative controls l Technical Specifications in the letter from Mr. William T.

Russell to Owner Group Technical Specification Committee -

Chairmen, dated October 25, 1993. According to Mr. Russell's letter, appropriate administrative Technical Specifications are those requirements not covered by other technical specification i sections, but are nacossary to assure operation of the plant in a safe manner.

The letter further presents five criteria for screening administrative controls: 1) requirements not covered by other regulatory requirements, but necessary to assure the safe ,

operation of the facility; 2) specific requirements that are broadly covered by regulations or other regulatory controls, for which details need to be specified in the Technical -

Specifications to ensure safe plant operation; 3) Technical Specifications should not duplicate other regulatory requirements; 4) specific reporting requirements should be retained that are not specified in regulations because they have been historically required by Technical Specifications; and 5) 10CFR50, Appendix I requirements under 50.36a. The specific reports defined in item 4 are: Occupational Radiation Exposure Report; Monthly Operating Report; and the Annual Radiological Environmental Operating Report (in the Prairie Island Technical Specifications and this License Amendment Request this report is known as the Annual Radiological Environmental Monitoring Report).

This License Amendment Request is framed within the guidance of Revision 1 to NUREG-1431 which in turn embodies the criteria defined in the previous paragraph for appropriate administrative controls.

)

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There are no plant modifications which are required to implement the changes requested in this License Amendment Request. Section 6 of the Technical Specifications addresses the administrative controls required to assure that Prairie Island is operated in a safe manner. These controls typically consist of programs and periodic reporting requirements. Revising the current administrative controls, removing current controls which duplicate requirements contained within the Code of Federal Regulations, or relocating control to other licensee programs does not require any plant modifications. Any changes to plant procedures, the Updated Safety Analysis Report, or the Operational Quality Assurance Plan which are necessary to support the changes requested by this License Amendment Request will also not result in any plant modifications.

The changes proposed by tnis License Amendment Request are administrative in nature and have not substantively revised any safety limits, limiting conditions for operation or surveillance requirements for the plant. Reformatting the current Administrative Controls section will provide a significant human factors improvement by locating similar requirements within the same section and provide a standard structure.

It is concluded that the administrative changes as requested by this amendment do not adversely affect public health and safety.

This amendment will, however, bring the Prairie Island Technical Specifications for plant procedures, programs, reports and other administrative controls closer to conformance with the industry standard.

DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS The proposed changes to the Operating License have been evaluated to 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 consequences of an accident previously evaluated Operation of the Prairie Island plant in accordance with the proposed changes does not involve a significant increase in the probability or consequences of an accident previously evaluated. None of the proposed changes involve a physical modification to the plant, a new mode of operation or a change to the Updated Safety Analysis Report transient analyses. These proposed amendments generally conform to the guidance of NUREG-1431, Revision 1, Section 5.0 which was previously reviewed, accepted and issued by the NRC.

Page 19

t l Some Section 5.0 Specifications in NUREG-1431 were not incorporated in this License Amendment Request. These Specifications were not proposed because they 1) specify requirements not currently in the Prairie Island Technical Specifications or otherwise committed to, 2)are addressed elsewhere in the current Technical Specifications, or 3) the current Technical Specifications level of commitment is maintained. In all these instances, the NRC has previously reviewed and approved the proposed level of commitment through the issuance of the current Prairie Island Technical Specifications.

The proposed changes, in themselves, do not reduce the level of qualification or training such that personnel requirements would be decreased.

In total these changes do 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 oreviousiv analyzed The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed changes, in themselves, do not introduce a new mode of plant operation, surveillance requirement or involve a physical modification to the plant.

These proposed amendments generally conform to the guidance of NUREG-1431, Revision 1, Section 5.0 which was previously reviewed, accepted and issued by the NRC.

Some Section 5.0 Specifications in NUREG-1431 were not incorporated in this License Amendment Request. These Specifications were not proposed because they 1) specify requirements not currently in the Prairie Island Technical Specifications or otherwise committed to, or 2) are addressed elsewhere in the current Technical Specifications. Other features are not fully implemented but rather, the current Technical Specification level of commitment is maintained. In all these instances, the NRC has previously reviewed and approved the' proposed level of commitment through the issuance of the current Prairie Island Technical Specitications.

In general, the proposed changes are administrative in nature. The changes propose to revise, delete or relocate Specifications within the Technical Specifications or from the Technical Specifications to the Updated Safety Analysis Report, plant procedures or the Operational Quality Assurance Plan through which adequate control is maintained. The Page 20

l proposed changes do not alter the design, function, or operation of any plant components and therefore, no new accident scenarios are created.

Therefore, the possibility of a new or different kind of accident from any accident previously evaluated would not .

be created be these amendments.

3. The proposed amendment will not involve a significant reduction in the marain of safety The proposed changes do not involve a significant reduction in a margin of safety because the current Technical Specifications requirements for safe operation of the Prairie Island plant are maintained or increased. The proposed changes are administrative in nature and do not involve a physical modification to the plant, a new mode of operation or a change to the Updated Safety Analysis Report transient analyses. The proposed changes do not alter the scope of equipment currently required to be operable or subject to surveillance testing nor does the proposed change affect any instrument setpoints or equipment safety functions.

Therefore, a significant reduction in the margin of safety would not be involved with these amendments.

Based on the evaluation described above, and pursuant to 110 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation the Prairie Island Nuclear Generating i

Plant in accordance with the proposed license amendment request does not involve any significant hazards considerations as defined by Nuclear Regulatory Commission 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 i

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.

I Page 21 l

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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