LIC-02-0105, License Amendment Request, Various Administrative and Editorial Changes

License Amendment Request, Various Administrative and Editorial Changes

ML030290240
Person / Time
Site:	Fort Calhoun
Issue date:	01/27/2003
From:	Bannister D Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-02-0105
Download: ML030290240 (26)

Text

toNh15fl Omaha Public Power Distnct 444 South 16th Street Mall Omaha NE 68102-2247 January 27, 2003 LIC-02-0105 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

References:

1.

Docket No. 50-285

SUBJECT:

Fort Calhoun Station Unit No. 1 License Amendment Request, "Various Administrative and Editorial Changes" Pursuant to 10 CFR 50.90, Omaha Public Power District (OPPD) hereby proposes to make administrative and editorial changes to the Fort Calhoun Station (FCS) Technical Specifications.

The proposed change consists primarily of editorial and typographical changes or corrections.

The proposed Technical Specification change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c); it has been determined that this change involves no significant hazards considerations.

The bases for these determinations, information supporting the change, a no significant hazards consideration, and an environmental consideration are included in the attached submittal.

Attachment I provides the Description of Changes and Justification. Attachment 2 provides the No Significant Hazards Evaluation and the technical bases for this requested change to the Technical Specifications. contains the marked-up pages and Attachment 4 contains the clean version reflecting the requested Technical Specification and Basis changes.

OPPD requests approval of the proposed amendment by June 30, 2003, to support scheduling implementation before the next refueling outage. OPPD requests that the effective date for this Technical Specification change be 60 days from issuance to allow for implementation of these proposed changes. No new commitments are made to the NRC in this letter.

Employment with Equal Opportunity 4171

U. S. Nuclear Regulatory Commission LIC-02-0105 Page 2 In accordance with 10 CFR 50.91, a copy of this application is being submitted to the designated Nebraska State Official.

I declare under penalty of perjury that the foregoing is true and correct. (Executed on January 27, 2003)

If you have any questions or require additional information, please contact Dr. R. L. Jaworski at (402) 533-6833.

Sincerely, D. J. Bannister Plant Manager Fort Calhoun Station DJB/RRL/rrl Attachments:

1. Description of Changes and Justification

2. Fort Calhoun Station's Evaluation

3. Markup of Technical Specification Pages

4. Proposed Technical Specifications (clean) c:

E. W. Merschoff, NRC Regional Administrator, Region IV A. B. Wang, NRC Project Manager J. G. Kramer, NRC Senior Resident Inspector Division Administrator - Public Health Assurance, State of Nebraska Winston & Strawn

LIC-02-0105 Page 1 Description of Changes and Justification CHANGE SPECIFICATION DESCRIPTION OF CHANGE AND JUSTIFICATION 1

1.3 Delete the last sentence of the second paragraph; "Provisions Basis (1) have been made to select different high-power level trip Page 1-6 points for various combinations of reactor coolant pump operation as described below under "Low Reactor Coolant Flow."

Justification: This sentence should have been removed when changes were made to support Amendment 92, but was overlooked. With Amendment 92, Fort Calhoun Station License was limited to 4-pump operation and removed the limiting safety system settings for two and three pump operation.

Amendment 92 negated the need to select different high-power level trip points for various combinations of reactor coolant pump operation. The "Low Reactor Coolant Flow" section, referred to in the sentence to be deleted by this proposed change, in basis Section 1.3(2) was properly corrected by Amendment 92.

Thus, the sentence proposed for deletion is meaningless and confusing. This proposed change is justified by the NRC Safety Evaluation Report issued with Amendment 92.

Therefore, this proposed change is administrative. [Example (i)]*

2 2.7(1)a Change "TIA-I" to "T1A-I" Page 2-32 Justification: This is a typographical error probably caused by a font that was not easily discemable as to whether the character was a 1 (one) or the capital letter "I" made some time ago and not caught by proofreading due to the similarity of the font. The tag number for the transformer is TI - T for transformer and 1 (one) for the first transformer.

The transformers specified in Section 2.7(1)a are correctly identified in Section 2.7(2)a. Therefore, this proposed change is administrative. [Example (i)]

3 2.7(1)b Page 2-32 Change "TIA-3" to "T1A-3" Justification: This is a typographical error probably caused by a font that was not easily discernable as to weather the character was a 1 (one) or the capital letter "I" made some

LIC-02-0105 Page 2 CHANGE SPECIFICATION DESCRIPTION OF CHANGE AND JUSTIFICATION time ago and not caught by proofreading due to the similarity of the font. The tag number for the transformer is T1. - T for transformer and 1 (one) for the first transformer.

The transformers specified in Section 2.7(1)b are correctly identified in Section 2.7(2)b and c. Therefore, this proposed change is administrative. [Example (i)]

4 2.7(1)d Change "TIB-3A, TIB3B, TIB-3C, TIB-4A, TIB-4B, TIB Page 2-32 4C" to "T1B-3A, T1B3B, T1B-3C, T1B-4A, T1B-4B, TIB 4C" Justification: This is a typographical error probably caused by a font that was not easily discernable as to whether the character was a 1 (one) or the capital letter "I" made some time ago and not caught by proofreading due to the similarity of the font. The tag number for the transformer is Ti. - T for transformer and 1 (one) for the first transformer.

The transformers specified in Section 2.7(1)d are correctly identified in Section 2.7(2)e.

Therefore, this proposed change is administrative. [Example (i)]

5 2.7(1)i.

Change "AI-40-A" to AI-40A" Page 2-32 Justification: This is a typographical error probably caused when the hyphen or dash was used to separate the tag "AI-40 from the "A, B, C and D," representing the four different Al 40 panels. The tag number for the instrument bus panels are "AI-40A, AI-40B, AI-40C, and AI-40D. Presently the dash or hyphen causes confusion, and should be removed.

Therefore, this proposed change is administrative. [Example (i)]

6 2.7 In the second full paragraph (the next-to-the-last and last Basis sentences) and in the third paragraph in this Technical Page 2-36a Specification Basis section, the term "safety-related" is being inserted to clarify these statements. That is, the phrase "the a-c instrument buses" is being changed to "the safety related a-c instrument buses."

Justification: Technical Specification Section 2.7 Basis requires changing to clarify which a-c instrument buses are actually being referred to. There are only four (4) safety

LIC-02-0105 Page 3 CHANGE SPECIFICATION DEScRIPTION OF CHANGE AND JUSTIFICATION related a-c instrument buses and un-interruptible power supplies (UPS) each of which feeds the four safety-related buses and the respective reactor protective system (RPS) channels.

However, in 1984 two additional, non-safety related un-interruptible power supplies (UPS) were added to feed the non-safety related instrument buses and the non safety related loads were removed from the safety related buses. The non-safety related a-c instrument buses or UPS do not power the RPS. At the time of this modification it was felt that this basis did not need to be changed.

However, over time, personnel have forgotten that Technical Specification 2.7 deals with the reactor protection system and engineered safeguards systems and their power supplies and are now confused as to whether this paragraph invokes requirements upon the non-safety related instrument buses.

This proposed change would minimize that confusion.

Therefore, this proposed change is administrative. [Example (i)]

7 3.0.2 Change the frequency for "Shift" from "At least once per 8 Page 3-0a hours," to "At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

Justification: This specification defines the frequency, or periodicity, at which "channel checks," operator log readings for temperature, level, pressure, etc. are taken, and equipment status is verified.

At the time that these specifications were written, the operations personnel were on an eight-hour shift. Operations shift personnel are now on a twelve-hour shift.

NUREG-1432, the improved standard technical specifications, does not provide a tabular format of the various surveillance frequency intervals as shown in 3.0.2.

NUREG-1432 (both Rev. 1 and 2), however, does prescribe that "channel checks," verification of level, position, mode, etc., and other required readings be performed at twelve hour intervals and, in some cases 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These are the same surveillances that would be affected by this change, that is, those surveillances listed in Table 3-1, 3-2, and 3-3 with a frequency requirement of"S," or once per shift.

This change is supported by the fact that operating experience has demonstrated the rarity of a channel failure and the probability of two random failures in redundant

LIC-02-0105 Page 4 CHANGE [SPECIFICATION IDESCRIPTION OF CHANGE AND JUSTIFICATION channels in any 12-hour period is extremely low. Many of these parameters are also displayed, alarmed, and logged on the plant computer giving the operator further warning of abnormal conditions. A catastrophic failure of the equipment subject to these surveillances would, in most cases, result in a plant alarm and not be discovered by the performance of the surveillance.

Thus, this change does not involve a change in the probability or consequences of accidents, or the possibility of a new or different type of accident. These surveillances serve primarily to minimize the chances of loss of protective function due to a slow degradation of equipment and to identify trends or subtle differences between redundant channels.

These surveillances supplement less formal, but more frequent, checks of channel operability, tank levels, system temperatures, etc.

made in the course of normal operational use of plant displays for the parameters associated with these surveillances.

The results of this slight change are clearly within all acceptance criteria with respect to the system or components specified in the Standard Review Plan. This is demonstrated by the fact that the NRC has previously approved the use of 12-hour surveillances at other U.S. commercial nuclear power plants and has found no adverse effects on plant safety resulting from this periodicity.

In that Monticello (Amendment 104, October 12, 1998, Accession number 9901050366 - 981244) was granted these changes and NUREG-1432, Rev. 1 and Rev. 2, were reviewed and approved by the NRC and several plants have converted to these specifications, this proposed change is administrative. [Example (vi & vii)]

8 Table 3-5 Revise the title of 11 so that it reads "Containment Item 11 Ventilation System Fusible Linked Dampers" Page 3-20d Justification: This change is needed to more accurately describe the subject fusible links and clarify the description.

With Amendment 198, Fort Calhoun Station changed the credit for containment iodine removal from the VA-3A and VA-3B fans to the Containment Spray System. This description should have been revised when changes were made to support Amendment 198, but was overlooked. This

LIC-02-0105 Page 5

• - The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14864) of amendments that are considered not likely to involve significant hazards consideration. Each of the above proposed changes cites one or more of these NRC provided examples. Example (i) relates to a purely administrative change to Technical Specifications: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature. Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications:

for example, a more stringent surveillance requirement. Example (iv) relates to a relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated. This assumes that the operating restriction and the criteria to be applied to a request for relief have been established in a prior review and that it is justified in a satisfactory way that the criteria have been met. Example (vi) relates to a change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptance criteria with respect to the system or component specified in the Standard Review Plan: for example, a change resulting from the application of a small refinement of a previously used calculational model or design method.

Example (vii) relates to a change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.

CHANGE SPECIFICATION DESCRIPTION OF CHANGE AND JUSTIFICATION proposed change is justified by the NRC Safety Evaluation Report issued with Amendment 198.

Therefore, this proposed change is administrative. [Example (i)]

9 3.5(3)ii Change "Personnel Access Lock" to "Personnel Air Lock" Page 3-37 Justification:

This change is being made to assure consistent use of the same terminology throughout the Technical Specifications.

Section 5.19(1) refers to "Personnel Air Lock" and this is also the commonly accepted terminology used in the FCS USAR, plant procedures, and by operations personnel.

Therefore, this proposed change is administrative. [Example (i)]

LIC-02-0105 Page 1 ATTACHMENT 2 Fort Calhoun Station's Evaluation for Amendment of Operating License

2.0 DESCRIPTION

OF PROPOSED AMENDMENT

3.0 BACKGROUND

4.0 REGULATORY REQUIREMENTS & GUIDANCE

6.0 REGULATORY ANALYSIS

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)

8.0 ENVIRONMENTAL CONSIDERATION

9.0 PRECEDENCE

10.0 REFERENCES

LIC-02-0105 Page 2 Fort Calhoun Station's Evaluation for Amendment of Operating License

1.0 INTRODUCTION

This letter is a request to amend Operating License DPR-40 for Fort Calhoun Station Unit No. 1.

The proposed amendment proposes to make administrative and editorial changes to the Fort Calhoun Station (FCS) Technical Specifications.

The proposed changes consist primarily of editorial and typographical changes or corrections.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT Fort Calhoun Station (FCS) proposes to change the Technical Specification (TS) conditions, specifications, requirements, and bases as described in Attachment 1, "Description of Changes and Justification."

The Technical Specifications affected are listed in the "Specification" column of Attachment 1. The "Description of Changes and Justification" column of Attachment 1 provides the proposed change and a justification for each change.

3.0 BACKGROUND

The proposed changes are a result of the Updated Safety Analysis Report (USAR)

Verification Project and the correction of discrepancies or confusing statements identified by the FCS condition report system.

4.0 REGULATORY REQUIREMENTS & GUIDANCE The proposed Technical Specifications changes are administrative in nature, primarily, typographical or editorial in nature. The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (Reference 10.2) of amendments that are considered not likely to involve significant hazards consideration. The "Description of Changes and Justification," Attachment 1, for each of the proposed changes cites one or more of these NRC provided examples.

5.0 TECHNICAL ANALYSIS

The proposed Technical Specifications changes are administrative in nature. Attachment 1, "Description of Changes and Justification" provides a description of each proposed

LIC-02-0105 Page 3 change and a justification as to why this change is considered administrative. Due to the number of changes and the various sections and specifications affected each change and a justification for the change is provided in Attachment 1, "Description of Changes and Justification."

6.0 REGULATORY ANALYSIS

The proposed Technical Specifications changes are administrative in nature. Attachment 1, "Description of Changes and Justification" provides a description of each proposed change and a justification as to why this change is considered administrative.

It should also be noted that the proposed changes do not alter, degrade, or prevent actions described or assumed in any accident analysis. They will not change any assumptions previously made in evaluating radiological consequences or affect any fission product barriers, nor do they increase any challenges to safety systems. Therefore, the proposed change does not increase or have any impact on the consequences of events described and evaluated in Chapter 14 of the Fort Calhoun USAR.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION Omaha Public Power District has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The correction of typographical errors and clarification of specifications is not an initiator of any previously evaluated accident. The frequency or periodicity of performance of those surveillances affected by this change are not an initiator of any previously evaluated accident. The proposed changes will not prevent safety systems from performing their accident mitigation function as assumed in the safety analysis.

Therefore, this change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

LIC-02-0105 Page 4

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change only affects the technical specifications and does not involve a physical change to the plant. Modifications will not be made to existing components nor will any new or different types of equipment be installed. The proposed change corrects typographical errors, provides clarification as to applicable equipment and modifies the frequency of surveillances performed once per shift from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This change will not alter assumptions made in safety analysis and licensing bases.

Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change corrects typographical errors, provides clarification as to applicable equipment, and modifies the frequency of surveillances performed once per shift from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The decrease in frequency or periodicity of performance of these surveillances will also permit more efficient and more safely managed plant operations and can help reduce the risk associated with changing plant equipment or operating modes in order to obtain some of these readings.

Therefore, this technical specification change does not involve a significant reduction in the margin of safety.

Based on the above, Omaha Public Power District concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

8.0 ENVIRONMENTAL CONSIDERATION

Based on the above considerations, the proposed amendment does not involve and will not result in a condition which significantly alters the impact of Fort Calhoun Station on the environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Part 51.22(c)(9), and, pursuant to 10 CFR Part 51.22(b), no environmental assessment need be prepared.

LIC-02-0105 Page 5 9.0 PRECEDENTS Similar amendment requests that would support this proposed change are cited in, Description of Changes and Justification, where required to support the fact that these changes are administrative changes.

10.0 REFERENCES

10.1 NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants" 10.2 Standards for Determining Whether License Amendments Involve No Significant Hazards Considerations, 48 FR 14864

LIC-02-0105 Page 1 ATTACHMENT 3 Markup of Technical Specification Pages

TECHNICAL SPECIFICATIONS 1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 Limiting Safety System Settings, Reactor Protective System Applicability This specification applies to RPS Limiting Safety System settings and bypasses for instrument channels.

Obiective To provide for automatic protection action in the event that the principal process variables approach a safety limit.

Specification The reactor protective system trip setting limits and the permissible bypasses for the instrument channels shall be within the Limiting Safety System Setting as stated in Table 1-1.

Basis The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached.

(1)

High Power Level - A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding resulting from some reactivity excursions too rapid to be detected by pressure and temperature measurements (in addition, thermal signals are provided to the high power level trip unit as a backup to the neutron flux signal).

During normal plant operation, with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 109.0% of indicated full power. Adding to this the possible variation in trip point due to calibration and measurement errors, the maximum actual steady-state power at which a trip would be actuated is 112%, which was used for the purpose of safety analysis.(1) Provisions have been made to selrt different high power level trip pointS for varGiou combinationS of reaGt*oF colant PUmP opeFation as described belnow under

--"Low React*oFr C-!oFt Flow!(-, )

During reactor operation at power levels between 19.1% and 100% of rated power, the Variable High Power Trip (VHPT) will initiate a reactor trip in the event of a reactivity excursion that increases reactor power by 10% or less of rated power. The high power trip setpoint can be set no more than 10% of rated power above the indicated plant power. Operator action is required to increase the set point as plant power is increased. The setpoint is automatically decreased as power decreases.

1-6 Amendment No. 5,32,210t

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.7 Electrical Systems Applicability Applies to the availability of electrical power for the operation of plant components.

Obiective To define those conditions of electrical power availability necessary to provide for safe reactor operation and the continuing availability of engineered safety features.

Specifications (1)

Minimum Requirements The reactor shall not be heated up or maintained at temperatures above 300OF unless the following electrical systems are operable:

a.

Unit auxiliary power transformers TILTA-1 or -2 (4,160 V).

b.

House service transformers T4,A-3 and 4 (4,160 V).

c.

4,160 V engineered safety feature buses 1A3 and 1A4.

d.

4,160 V/480 V Transformers TIIFB-3A, TI-fB-3B, TffB-3C, TITB-4A, TIt1B-4B, TtB-4C.

e.

480 V distribution buses 1B3A, 1B3A-4A, 1B4A, 1B3B, 1B3B-4B, 1B4B, 1B3C, 1B3C-4C, 1B4C.

f.

MCC No. 3A1, 3B1, 3A2, 3C1, 3C2, 4A1, 4A2, 4C1 and 4C2.

g.

125 V d-c buses No. 1 and 2 (Panels EE-8F and EE-8G).

h.

125 V d-c distribution panels AI-41A and AI-41 B.

i.

120V a-c instrument buses A, B, C, and D (Panels AI-40-A, B, C and D).

j.

Inverters A, B, C, and D.

k.

Station batteries No. 1 and 2 (EE-8A and EE-8B) including one battery charger on each 125V d-c bus No. 1 and 2 (EE-8F and EE-8G).

I.

Two emergency diesel generators (DG-1 and DG-2).

m.

One diesel fuel storage system containing a minimum volume of 16,000 gallons of diesel fuel in FO-1, and an additional 8,000 gallons of diesel fuel in FO-10.

2-32 Amendment No. 447,462,180

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.7 Electrical Systems (Continued) plant shutdown as required by Specification 2.7(2). This period is acceptable based on the remaining capacity (more than 6 days), the fact that procedures are in place to obtain replenishment, and the low probability of an event during this brief period.

Additional supplies of diesel fuel oil are available in the Omaha area and from nearby terminals. Ample facilities exist to assure deliveries to the site within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

One battery charger on each battery shall be operating so that the batteries will always be at full charge; this ensures that adequate d-c power will be available for all emergency uses. Each battery has one battery charger permanently connected with a third charger capable of being connected to either battery bus. The chargers are each rated for 400 amperes at 130 volts. Following a DBA the batteries and the chargers will handle all required loads. Each of the reactor protective channels instrumentation channels is supplied by one of the F

-Ela:tde a-c instrument buses. The removal of one of the ýaf6-+e6tqd a-c instrument buses is permitted as the 2-of-4 logic may be manually changed to a 2-of-3 logic without compromising safety.

The engineered safeguards instrument channels use Kat.-iLat c a-c instrument buses (one redundant bus for each channel) and d-c buses (one redundant bus for each logic circuit). The removal of one of the §FRO-j a-c instrument buses is permitted as the two of four logic automatically becomes a two of three logic.

Required engineered safeguards components, as described in Specification 2.7(2),

refers to components required to be operable by other Limiting Conditions for Operation within these Technical Specifications. If no other LCO requires a particular ESF component to be operable, then its redundant component is also not required to be operable due to this specification. As an example, Specification 2.3 requires that safety injection pumps be operable prior to the reactor being made critical, and Specification 2.7 applies when the RCS is above 300EF. If the RCS is above 300EF but the reactor is not critical, then no safety injection pumps are required to be operable.

References (1) USAR, Section 8.3.1.2 (2) USAR, Section 8.4.1 (3) USAR, Section 8.2.2 Amendment No. 147,162,180-2-36a

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.0.1 Each surveillance requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

3.0.2 The surveillance intervals are defined as follows:

Notation Title Frequency S

Shift At least once per 8C2 hours D

Daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W

Weekly At least once per 7 days BW Biweekly At least once per 14 days M

Monthly At least once per 31 days Q

Quarterly At least once per 92 days SA Semiannual At least once per 184 days A

Annually At least once per 366 days R

Refueling At least once per 18 months P

Start up Prior to Reactor Start up, if not completed in the previous week.

Exception to these intervals are stated in the individual Specifications.

3.0.3 The provisions of Specifications 3.0.1 and 3.0.2 are applicable to all codes and standards referenced within the Technical Specifications. The requirements of the Technical Specifications shall have precedence over the requirements of the codes and standards referenced within the Technical Specifications.

3.0.4 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specifications 3.0.1 and 3.0.2, shall constitute noncompliance with the OPERABILITY requirements for the corresponding Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

Amendment No. 122,129,57-3-0a

TECHNICAL SPECIFICATIONS Test TABLE 3-5 (Continued)

Frequency USAR Section Reference 1Oc.

(continued)

11.

Containment Goal ing-and-ledine Removal Ventilation System Fusible Linked Dampers

12.

Diesel Generator Under-Voltage Relays

13.

Motor Operated Safety Injection Loop Valve Motor Starters (HCV-31 1, 314, 317, 320,327, 329, 331, 333, 312, 315, 318, 321)

14.

Pressurizer Heaters

15.

Spent Fuel Pool Racks

16.

Reactor Coolant Gas Vent System

4. Automatic and/or manual initi ation of the system shall be demonstrated.

1. Demonstrate damper action.

2. Test a spare fusible link.

Calibrate Verify the contactor pickup value at

<85% of 460 V.

Verify control circuits operation for post-accident heater use.

Test neutron poison samples for dimensional change, weight, neutron attenuation change and specific gravity change.

1. Verify all manual isolation valves in each vent path are in the open position.

2. Cycle each automatic valve in the vent path through at least one complete cycle of full travel from the control room. Verification of valve cycling may be determined by observation of position indicating lights.

3. Verify flow through the reactor coolant vent system vent paths.

R 1 year, 2 years, 5 years, and every 5 years thereafter.

R R

R 1, 2, 4, 7, and 10 years after installation, and every 5 years thereafter.

During each refueling outage just prior to plant start-up.

R R

Amendment No. 41,54,60,75,77,80,155,169,-82 9.10 8.4.3 3-20d

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Test Applicability Applies to containment leakage and structural integrity.

Obiective To verify that the:

(1)

Locked closed manual containment isolation valves are closed and locked, (2) potential leakage from containment is within acceptable limits, and (3) structural performance of all important components in the containment prestressing system is acceptable.

Specifications (1)

Prior to the reactor going critical after a refueling outage, and at least once per 31 days thereafter, an administrative check will be made to confirm that all "locked closed" manual containment isolation valves, except for valves that are open under administrative control as permitted by Specification 2.6(1)a, are closed and locked.

Valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position shall be verified closed during each cold shutdown except that such verification need not be performed more often than once per 92 days.

(2)

Containment Integrated Leakage Rate Test (Type A Tests)

Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program.

(3)

Containment Penetrations Leak Rate Tests (Type B Tests)

Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program for the following penetrations:

(i)

Equipment Hatch (ii)

Personnel AGoess KJf Lock (iii)

Mechanical Penetrations M-1 through M-99 (iv)

Fuel Transfer Tube (Mechanical Penetration M-100)

(v)

Electrical Penetrations:

A-1 B-9 D-6 F-2 E-HCV-383-3A A-2 B-10 D-7 F-4 E-HCV-383-3B A-4 B-11 D-8 F-5 E-HCV-383-4A A-5 C-1 D-9 F-6 E-HCV-383-4B A-6 C-2 D-10 F-7 A-7 C-4 D-11 F-8 A-8 C-5 E-1 F-9 A-9 C-6 E-2 F-1 0 Amendment No. 95,!51,185 3-37

LIC-02-0105 Page 1 ATTACHMENT 4 Proposed Technical Specification Pages

TECHNICAL SPECIFICATIONS 1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 Limiting Safety System Settings, Reactor Protective System Applicability This specification applies to RPS Limiting Safety System settings and bypasses for instrument channels.

Obiective To provide for automatic protection action in the event that the principal process variables approach a safety limit.

Specification The reactor protective system trip setting limits and the permissible bypasses for the instrument channels shall be within the Limiting Safety System Setting as stated in Table 1-1.

Basis The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached.

(1)

High Power Level - A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding resulting from some reactivity excursions too rapid to be detected by pressure and temperature measurements (in addition, thermal signals are provided to the high power level trip unit as a backup to the neutron flux signal).

During normal plant operation, with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 109.0% of indicated full power. Adding to this the possible variation in trip point due to calibration and measurement errors, the maximum actual steady-state power at which a trip would be actuated is 112%, which was used for the purpose of safety analysis.(1)

During reactor operation at power levels between 19.1% and 100% of rated power, the Variable High Power Trip (VHPT) will initiate a reactor trip in the event of a reactivity excursion that increases reactor power by 10% or less of rated power. The high power trip setpoint can be set no more than 10% of rated power above the indicated plant power. Operator action is required to increase the set point as plant power is increased. The setpoint is automatically decreased as power decreases.

Amendment No. 5,32,21-0 1-6

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.7 Electrical Systems Applicability Applies to the availability of electrical power for the operation of plant components.

Obiective To define those conditions of electrical power availability necessary to provide for safe reactor operation and the continuing availability of engineered safety features.

Specifications (1)

Minimum Requirements The reactor shall not be heated up or maintained at temperatures above 300OF unless the following electrical systems are operable:

a.

Unit auxiliary power transformers T1A-1 or -2 (4,160 V).

b.

House service transformers T1A-3 and 4 (4,160 V).

c.

4,160 V engineered safety feature buses 1A3 and 1A4.

d.

4,160 V/480 V Transformers T1 B-3A, T1B-3B, T1B-3C, T1B-4A, T1B-4B, TI B-4C.

e.

480 V distribution buses 1B3A, 1B3A-4A, 1B4A, 1B3B, 1B3B-4B, 1B4B, 1B3C, 1 B3C-4C, 1B4C.

f.

MCC No. 3A1, 3B1, 3A2, 3C1, 3C2, 4A1, 4A2, 4C1 and 4C2.

g.

125 V d-c buses No. I and 2 (Panels EE-8F and EE-8G).

h.

125 V d-c distribution panels AI-41A and AI-41B.

i.

120V a-c instrument buses A, B, C, and D (Panels AI-40A, B, C and D).

j.

Inverters A, B, C, and D.

k.

Station batteries No. 1 and 2 (EE-8A and EE-8B) including one battery charger on each 125V d-c bus No. 1 and 2 (EE-8F and EE-8G).

I.

Two emergency diesel generators (DG-1 and DG-2).

m.

One diesel fuel storage system containing a minimum volume of 16,000 gallons of diesel fuel in FO-1, and an additional 8,000 gallons of diesel fuel in FO-10.

Amendment No. 147,162,180 2-32

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.7 Electrical Systems (Continued) plant shutdown as required by Specification 2.7(2). This period is acceptable based on the remaining capacity (more than 6 days), the fact that procedures are in place to obtain replenishment, and the low probability of an event during this brief period.

Additional supplies of diesel fuel oil are available in the Omaha area and from nearby terminals. Ample facilities exist to assure deliveries to the site within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

One battery charger on each battery shall be operating so that the batteries will always be at full charge; this ensures that adequate d-c power will be available for all emergency uses. Each battery has one battery charger permanently connected with a third charger capable of being connected to either battery bus. The chargers are each rated for 400 amperes at 130 volts. Following a DBA the batteries and the chargers will handle all required loads. Each of the reactor protective channels instrumentation channels is supplied by one of the safety-related a-c instrument buses. The removal of one of the safety-related a-c instrument buses is permitted as the 2-of-4 logic may be manually changed to a 2-of-3 logic without compromising safety.

The engineered safeguards instrument channels use safety-related a-c instrument buses (one redundant bus for each channel) and d-c buses (one redundant bus for each logic circuit). The removal of one of the safety-related a-c instrument buses is permitted as the two of four logic automatically becomes a two of three logic.

Required engineered safeguards components, as described in Specification 2.7(2),

refers to components required to be operable by other Limiting Conditions for Operation within these Technical Specifications. If no other LCO requires a particular ESF component to be operable, then its redundant component is also not required to be operable due to this specification. As an example, Specification 2.3 requires that safety injection pumps be operable prior to the reactor being made critical, and Specification 2.7 applies when the RCS is above 300EF. If the RCS is above 300EF but the reactor is not critical, then no safety injection pumps are required to be operable.

References (1) USAR, Section 8.3.1.2 (2) USAR, Section 8.4.1 (3) USAR, Section 8.2.2 Amendment No. 147,!62,180 2-36a

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.0.1 Each surveillance requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

3.0.2 The surveillance intervals are defined as follows:

Notation Title Frequency S

Shift At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D

Daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W

Weekly At least once per 7 days BW Biweekly At least once per 14 days M

Monthly At least once per 31 days Q

Quarterly At least once per 92 days SA Semiannual At least once per 184 days A

Annually At least once per 366 days R

Refueling At least once per 18 months P

Start up Prior to Reactor Start up, if not completed in the previous week.

Exception to these intervals are stated in the individual Specifications.

3.0.3 The provisions of Specifications 3.0.1 and 3.0.2 are applicable to all codes and standards referenced within the Technical Specifications. The requirements of the Technical Specifications shall have precedence over the requirements of the codes and standards referenced within the Technical Specifications.

3.0.4 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specifications 3.0.1 and 3.0.2, shall constitute noncompliance with the OPERABILITY requirements for the corresponding Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

Amendment No. 422,1o29,*157-,

3-0a

TECHNICAL SPECIFICATIONS Test TABLE 3-5 (Continued)

Frequency USAR Section Reference 1Oc.

(continued)

11.

Containment Vent lation System Fusible Linked Dampers

12.

Diesel Generator Under-Voltage Relays

13.

Motor Operated Safety Injection Loop Valve Motor Starters (HCV-31 1, 314, 317, 320, 327, 329, 331, 333, 312, 315, 318, 321)

14.

Pressurizer Heaters

15.

Spent Fuel Pool Racks

16.

Reactor Coolant Gas Vent System

4. Automatic and/or manual initi ation of the system shall be demonstrated.

1. Demonstrate damper action.

2. Test a spare fusible link.

Calibrate Verify the contactor pickup value at

<85% of 460 V.

Verify control circuits operation for post-accident heater use.

Test neutron poison samples for dimensional change, weight, neutron attenuation change and specific gravity change.

1. Verify all manual isolation valves in each vent path are in the open position.

2. Cycle each automatic valve in the vent path through at least one complete cycle of full travel from the control room. Verification of valve cycling may be determined by observation of position indicating lights.

3. Verify flow through the reactor coolant vent system vent paths.

R 1 year, 2 years, 5 years, and every 5 years thereafter.

R R

R 1, 2, 4, 7, and 10 years after installation, and every 5 years thereafter.

During each refueling outage just prior to plant start-up.

R R

Amendment No. 41,54,60,75,77,80,155,169,482 9.10 8.4.3 3-20d

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Test Applicability Applies to containment leakage and structural integrity.

Obiective To verify that the:

(1)

Locked closed manual containment isolation valves are closed and locked, (2) potential leakage from containment is within acceptable limits, and (3) structural performance of all important components in the containment prestressing system is acceptable.

Specifications (1)

Prior to the reactor going critical after a refueling outage, and at least once per 31 days thereafter, an administrative check will be made to confirm that all "locked closed" manual containment isolation valves, except for valves that are open under administrative control as permitted by Specification 2.6(1)a, are closed and locked.

Valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position shall be verified closed during each cold shutdown except that such verification need not be performed more often than once per 92 days.

(2)

Containment Integrated Leakage Rate Test (Type A Tests)

Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program.

(3)

Containment Penetrations Leak Rate Tests (Type B Tests)

Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program for the following penetrations:

(i)

Equipment Hatch (ii)

Personnel Air Lock (iii)

Mechanical Penetrations M-1 through M-99 (iv)

Fuel Transfer Tube (Mechanical Penetration M-100)

(v)

Electrical Penetrations:

A-1 B-9 D-6 F-2 E-HCV-383-3A A-2 B-10 D-7 F-4 E-HCV-383-3B A-4 B-1 1 D-8 F-5 E-HCV-383-4A A-5 C-1 D-9 F-6 E-HCV-383-4B A-6 C-2 D-10 F-7 A-7 C-4 D-11 F-8 A-8 C-5 E-1 F-9 A-9 C-6 E-2 F-1 0 Amendment No. 95,451,-85 3-37