LIC-02-0130, License Amendment Request, Pressurizer Safety Valve Lift Settings & Administrative Revisions.

License Amendment Request, Pressurizer Safety Valve Lift Settings & Administrative Revisions.

ML030290219
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-02-0130
Download: ML030290219 (23)

Text

Opp" m m Omaha PubicPower Distrct 444 South 16th Street Mall Omaha NE 68102-2247 January'27, 2003 LIC-02-0130 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Reference:

Docket No. 50-285

SUBJECT:

Fort Calhoun Station Unit No. 1 License Amendment Request, "Pressurizer Safety Valve (PSV) Lift Settings and Administrative Revisions" Pursuant to 10 CFR 50.90, Omaha Public Power District (OPPD) requests to amend Technical Specifications (TS) 2.1.6, 3.2 (Table 3-5), and TS 5.9.1c as follows. In TS 2.1.6(1), the "as found" PSV lift setting tolerance band of +/- 1% is increased to +1%/ -3% to allow for normal setpoint variance for Modes 1 and 2. The Basis of TS 2.1.6 is revised to clarify that the PSVs are still operable and capable of performing their safety function with the wider tolerance band. The remaining revisions to TS 2.1.6 are administrative in nature to change defined terms to upper case text.

TS 3.2, Table 3-5, Item 3 is revised to require an "as-left" PSV lift setting tolerance band of-4 1%. TS 5.9.1c is revised to remove the requirement to provide a statement in the Monthly Operating Report (MOR) concerning failures or challenges to Power Operated Relief Valves (PORV) or Safety Valves. Generic Letter 97-02, Revised Contents of the Monthly Operating Report does not require the MOR to provide this information.

Attachment 1 provides the No Significant Hazards Evaluation and the technical bases for the changes requested to the Technical Specifications. Attachments 2 and 3 contain a marked-up and clean version reflecting the requested Technical Specification and Basis changes.

OPPD requests approval of the proposed amendment by August 1, 2003 to support scheduling its implementation before the 2003 Refueling Outage that begins September 12, 2003. OPPD requests that the effective date for this Technical Specification change be 30 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-02-0130 Page 2 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 of my staff at 402-533-6833.

Sincerely, D. J. Bannister Plant Manger Fort Calhoun Station DJB/RLJ/rlj Attachments

1. Fort Calhoun Station's Evaluation for Amendment of Operating License

2. Mark-up of Technical Specifications

3. Clean Version of Technical Specifications 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-02-0130 Page 1 ATTACHMENT 1 Fort Calhoun Station's Evaluation For Pressurizer Safety Valve (PSV) Lift Setting and Administrative Revisions

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-02-0130 Page 2

1.0 INTRODUCTION

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

The Omaha Public Power District (OPPD) proposes to change the lift setting tolerance band for operability of the pressurizer safety valves (PSVs) from +/- 1% to +1%/ -3% for Modes 1 and 2 to allow for normal setpoint variance. The proposed amendment specifies that "as-left" PSV lift setting tolerance bands of +/- 1% must be met at the conclusion of surveillance testing (ST). OPPD proposes to revise the Bases to summarize the information supporting the proposed +1%/ -3% tolerance band and delete non-value-added text.

OPPD proposes to change defined terms currently in lowercase text to uppercase text for consistency with Standard Technical Specification (STS). The requirement to provide a statement in the Monthly Operating Report (MOR) concerning failures or challenges to power operated relief valves (PORVs) or safety valves is also proposed for deletion.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT The Technical Specification (TS) 2.1.6(1) Limiting Conditions for Operation (LCO) for the PSV lift setting tolerance band is increased from +/- 1% to +1%/ -3%. This wider tolerance band allows for normal setpoint variance within which the valves will still perform their design basis safety function. OPPD is revising Table 3-5 to require "as-left" PSV lift settings of +/- 1% at the conclusion of surveillance testing. The Basis of TS 2.1.6 is being revised to summarize and update information supporting the proposed TS changes.

The remaining revisions to TS 2.1.6 are administrative in nature. Many, but not all defined terms in TS 2.1.6 are in uppercase text to conform to the STS format. Therefore, the remaining defined terms currently in lowercase text were changed to uppercase text for consistency with STS. Finally, OPPD proposes to revise TS 5.9.1c to remove the requirement to provide a statement in the MOR concerning failures or challenges to the PORVs or safety valves.

3.0 BACKGROUND

The unnecessarily restrictive =- 1% PSV lift setting tolerance band in TS 2.1.6(1) has resulted in the submittal of several non-safety-significant Licensee Event Reports (LERs) to report PSVs that opened outside this tolerance band during surveillance testing. In 1993, significant improvements to the FCS PSV testing methodology were implemented.

After testing in 1996, procedural guidance was clarified to ensure better temperature control. Nevertheless, three of the eight "as-found" PSV tests since 1996 were not within LIC-02-0130 Page 3

+/- 1% of the lift setting. In each case, it was determined that the condition was not safety significant, and neither the testing nor the valve itself was determined to be deficient.

The conclusion was that the out-of-tolerance test results were attributable to normal setpoint variance. This phenomenon is discussed in Section 5.4 of Electric Power Research Institute (EPRI) Report TR-105872, "Safety and Relief Valve Testing and Maintenance Guide," dated August 1996. The report recommends establishing a tolerance band of wider than +/-1% for tests of valves that have been in service, if justified by engineering review, while still requiring the valves to be set to within a +/- 1% tolerance band before concluding testing. Since the current requirement to set the valves to within +/-

1% before returning them to service would remain in effect, the anticipated performance of the valves over the course of the subsequent operating cycle would not change.

Standard Technical Specifications use capitalized type for defined terms in the Technical Specifications and Bases. A review of TS 2.1.6 found several non-capitalized defined terms. These non-capitalized defined terms were capitalized for consistency with STS and the capitalized defined terms in TS 2.1.6.

The FCS MOR currently requires a statement concerning failures or challenges to the PORVs or safety valves. Virtually all replies to this statement are "None." On those rare occasions when there is a challenge or failure, the statement typically refers the reader to an LER for details of the event. Thus, the MOR statement is redundant to LER notification.

Furthermore, Generic Letter (GL) 97-02, states in part, "Effective immediately, licensees of operating nuclear power plants submitting monthly operating reports called for in the Technical Specifications may do so in accordance with the guidance provided in Attachment 1 to this generic letter. ... However, licensees will have to take whatever means are appropriate to negate any prior commitments or requirements to provide monthly operating reports which contain the information identified in Draft Regulatory Guide 1.16, Revision 4, Section C.l.c; this may include an amendment to the facility operating license to remove a license condition." There is no requirement in Attachment 1 of GL 97-02 for a statement pertaining to challenges or failures of the PORVs or safety valves.

4.0 REGULATORY REQUIREMENTS & GUIDANCE The Fort Calhoun Station design meets criteria similar to those now contained in 10 CFR 50 Appendix A, General Design Criteria for Nuclear Power Plants. The Fort Calhoun Station USAR Appendix G, Responses to 70 Criteria, describes how those criteria are met.

The criteria that govern reactor coolant pressure are USAR Appendix G Criterion 6, 9, and 14.

Criterion 6 - Reactor Core Design states: "The core design together with reliable process and decay heat removal systems, shall provide for this capability under all expected con-LIC-02-0130 Page 4 ditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine-generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power." "In an instance of a turbine trip, the reactor is scrammed and energy transfer to the secondary systems continues by means of automatic bypassing of steam directly to the main condenser. In addition, relief valves are provided to directly limit reactor coolant system pressure and temperature in the event of isolation of the reactor from its primary heat sink."

Criterion 9 - Reactor Coolant Pressure Boundary states: "Reactor coolant system components are designed for a pressure of 2500 psia and a temperature of 650°F. The nominal operating conditions of 2100 psia and an average reactor coolant system temperature of 572.5°F permit an adequate margin for normal load changes and operating transients. The components are designed and constructed in accordance with the ASME Boiler & Pressure Vessel Code, Section IlI, and as delineated in Criterion 1."

Section III of the ASME Boiler & Pressure Vessel Code states: "The rated capacity of the pressure relieving devices including any limitation imposed by the systems connected to the discharge side shall be sufficient to prevent a rise in pressure within the vessels which they protect of more than 10 percent above the design pressure at the design temperature when the pressure relieving devices are operating under the condition summarized in the technical report of N-910.2."

Criterion 14 - Core Protection Systems states: "The reactor is protected by the reactor protection system from reaching a condition at which fuel damage might occur. The protection system is designed to monitor the reactor operating conditions and initiate a fast shutdown if any of the measured variables exceed the operating limits. The signals, which will provide automatic reactor trip, are identified in Table 7.2-1 of the Fort Calhoun FSAR.

The parameters and conditions, which will initiate a trip, are the following: ...

c) High Pressurizer Pressure d) Thermal Margin/Low Pressure e) Loss of Load

5.0 TECHNICAL ANALYSIS

5.1 Design Basis Fort Calhoun Station has two spring-loaded PSVs that provide over-pressure protection for the Reactor Coolant System (RCS).

Together with the Reactor Protection System, the PSVs ensure that RCS pressure does not exceed the 2750 psia Safety Limit (SL). The PSVs have a water-filled loop seal at the inlet to each valve. As indicated in the current TS 2.1.6 Basis, the ASME Code requires steam to be used as the test medium for establishing the setpoint for valves in LIC-02-0130 Page 5 steam service. However, the presence of the loop seal may result in in-situ PSV actuation at a pressure that differs from the setpoint pressure established during testing with steam.

The safety analysis addresses this issue with a Loss of Load case that assumes actuation pressures up to 6% above the nameplate setpoint for both PSVs. RCS pressure was found to remain below the SL in this case even with this assumption.

In 1992, as a result of LER 92-028, FCS used test data, data from operating events, and an analytical model to evaluate the relationship between PSV lift pressures established during testing with steam (no loop seal) versus expected lift pressure for in-service PSVs with a water-filled loop seal at the valve inlet. Based on the evaluation (Reference 10.5),

an improved testing procedure was established. Evaluation of the data and analytical model indicated that a PSV that had its lift setting established on steam using the improved test procedure would initially lift at a pressure of up to 1% below this setting when installed in the plant on a loop seal.

Opening pressures below the specified setpoint are not a concern with respect to the safety limit for RCS pressure. The PSV setpoints are above the high pressurizer pressure reactor trip setpoint that opens the PORVs. Specifically, the lowest setpoint PSV (i.e.,

RC-142 with a nominal setpoint of 2485 psig, which correlates to about 2500 psia) is about 4% above the TS 1.2 upper limit (i.e., 2400 psia) for the reactor trip and PORV actuation setting. The nominal reactor trip/PORV actuation setting that is currently established by procedure is 2350 psia. As a result, there is about a 6% margin between the nominal reactor trip/PORV actuation setting and the nominal setpoint of RC-142.

This margin is sufficient to accommodate the combination of the TS 2.1.6(1) -3% PSV lift setting limit and the anticipated 1% reduction in actuation pressure due to the loop seal, without compromising RCS integrity during power operation. Therefore, a PSV is operable at a lift setting up to 3% below the nominal setpoint.

Since Table 3-5 is being changed to require a lift setting within +/- 1% after testing, the tolerance band that must be met before returning a PSV to service after testing is unchanged.

Therefore, the proposed amendment does not change the design basis or methodology for the PSV lift setting tolerance band. The remaining changes proposed for Sections 2.1.6 and 5.9.1.c., are administrative in nature to provide consistency of capitalization of defined terms and compliance with GL 97-02 and thus do not change the design basis or methodology.

5.2 Risk Information The proposed amendment does not involve application or use of risk-informed decisions.

The risk to the health and safety of the public because of the changes to PSV lift setting tolerance band, associated surveillance testing, Bases, and administrative revisions, is minimal.

LIC-02-0130 Page 6

6.0 REGULATORY ANALYSIS

The proposed amendment increases the "as-found" PSV lift setting tolerance band from +/-

1% to +1%! -3% to allow for normal setpoint variance. The proposed tolerance band does not compromise RCS integrity during power operation and does not adversely affect over-pressure protection. The proposed amendment clarifies that the PSV lift setting is within +/- 1% of the nominal setpoint at the conclusion of surveillance testing. The remaining changes provide support for the PSV lift setting tolerance band change, provide consistency of defined term usage, and revise information contained in the MOR in accordance with GL 97-02.

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 OPPD has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) 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 proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The design basis event for RCS over-pressure protection is the Loss of Load accident. The Loss of Load event was previously evaluated assuming the PSVs lift up to 6% above their setpoint. While the proposed amendment widens the tolerance band for installed PSVs, only the lower end of the band is changed; therefore, there is no adverse affect on the over-pressure protection analysis.

The proposed amendment does not change the tolerance band currently required at the conclusion of PSV surveillance testing each refueling outage. As with the current specification, the PSVs will continue to be set to within a tolerance band of+/-

1% using ASME Code test methods. As a result, the anticipated performance of the valves over the course of the subsequent operating cycle is not changed. In other words, the potential for setpoint variance exists regardless of whether the TSs are changed. The PSVs will begin each operating cycle after having been set to open LIC-02-0130 Page 7 within a lift setting tolerance band of +/- 1%. Therefore, the probability or consequences of potential setpoint variance during an operating cycle does not change. The remaining changes provide supporting statements for the wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are administrative in nature, or are in accordance with GL 97-02.

The changes in the case of the defined terms and elimination of the TS 5.9.1c Monthly Operating Report concerning failures or challenges to PORVs or safety valves are administrative changes which do not affect the initiator of an event or prevent safety systems from performing their accident mitigation functions as assumed in the safety analysis.

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

Response: No.

Widening the lift setting tolerance band for installed PSVs does not create the possibility of a new or different type of accident from any previously evaluated.

The accident analyses address the lift setting tolerance band of the PSVs, and the proposed tolerance band does not adversely affect the over-pressure protection function and will not compromise RCS integrity during power operation. No physical changes to the plant are involved.

The proposed amendment does not change the tolerance band that must be met at the conclusion of PSV surveillance testing each refueling outage. As with the current Technical Specifications, the PSVs will continue to be set at a tolerance band of +/-

1% using ASME Code test methods. As a result, the anticipated performance of the valves over the course of the subsequent operating cycle is not changed. The remaining changes provide supporting statements for the wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are administrative in nature, or are in accordance with GL 97-02 and thus do not create the possibility of a new or different type of accident from any previously evaluated.

The changes in the case of the defined terms and elimination of the TS 5.9.1c Monthly Operating Report concerning failures or challenges to PORVs or safety valves are administrative changes which only affect the technical specifications and do not involve a physical change to the plant. Therefore these changes do not alter assumptions made in the safety analysis and licensing basis.

LIC-02-0130 Page 8

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

Response: No.

Widening the lift setting tolerance band for installed PSVs does not involve a significant reduction in a margin of safety. The tolerance band of the PSVs is addressed in the accident analyses, and the proposed tolerance band does not adversely affect the over-pressure protection analysis. No physical changes to the plant are involved.

The proposed amendment does not change the tolerance band that must be met at the conclusion of PSV surveillance testing each refueling outage. As with the current Technical Specifications, the PSVs will continue to be set to a tolerance band of +/- 1%

using ASME Code test methods. As a result, the anticipated performance of the valves over the course of the subsequent operating cycle is not changed. The remaining changes provide supporting statements for the wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are administrative in nature, or are in accordance with GL 97-02.

The changes in the case of the defined terms and elimination of the TS 5.9.1c Monthly Operating Report concerning failures or challenges to PORVs or safety valves are administrative changes which only affect the technical specifications and reporting frequency.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above discussion, OPPD 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

The changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1) As demonstrated in Section 7.0, the proposed amendment does not involve a significant hazards consideration.

2) The proposed amendment does not result in a significant change in the types or increase in the amounts of any effluents that may be released offsite. In addition, the TS change does not introduce any new effluents or significantly increase the LIC-02-0130 Page 9 quantities of existing effluents. As such, the change cannot significantly affect the types or amounts of any effluents that may be released offsite.

3) The proposed amendment does not result in a significant increase in individual or cumulative occupational radiation exposure. The proposed change does not result in any physical plant changes. No new surveillance requirements are anticipated because of the changes that would require additional personnel entry into radiation controlled areas. Therefore, the amendment has no significant affect on either individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

9.0 PRECEDENCE Amendments allowing an increase in the as-found setpoint tolerance band for Pressurizer Safety Valves were approved for Virginia Power Company's North Anna Units No. 1 and 2 (Docket Nos. 50-338 and 50-339, dated April 1, 1996), Southern Nuclear Operating Company's Vogtle Electric Generating Plant, Units No. 1 and 2 (Docket Nos. 50-424 and 50-425, dated August 26, 1997), and Union Electric Company's Callaway Plant (Docket No. 50-483), dated May 25, 2000 (TAC No. MA9080).

10. REFERENCES 10.1 Fort Calhoun Unit 1 Technical Specifications Section Pressurizer and Main Steam Safety Valves 10.2 Fort Calhoun Unit 1 Technical Specifications Section 3.2, Equipment and Sampling Tests 10.3 Fort Calhoun Unit 1 Technical Specifications Section 5.9.1, Routine Reports 10.4 Generic Letter 97-02, "Revised Contents of the Monthly Operating Report," dated May 15, 1997 10.5 FCS Engineering Analysis No. EA-FC-92-067, "Pressurizer Safety Valves RC-141 & 142 Setpoint Evaluation 'With Loop-Seal'," Revision 0, September 1, 1992.

LIC-02-0130 Page 1 ATTACHMENT 2 Mark-up of Technical Specifications 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves Applicability Applies to the status of the pressurizer and main steam safety valves.

Objective To specify minimum requirements pertaining to the pressurizer and main steam safety valves.

Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:

(1) The reactor shal: not be made critical uinless the two presrie safety valves are operable with their loft settings adjusted to ensure valeopin at 2485 psig +/-1 % and "2530 psi--"-' Two pressurizer safety valves shall be OPERABLE in MODES 1 and 2, with lift settings of 2485 psig +1 %/-3% and 2530 psig +1 %/-3% respectively.

(2) Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable OPERABLE safety valve shall be installed on the pressurizer.

However, when in at least the cold shutdown COLD SHUTDOWN condition, safety valve nozzles may be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification.

(3) At least four of the five Main Steam Safety Valves (MSSVs) associated with each steam generator shall be OPERABLE in MODES 1 and 2. Lift settings shall be at 985 psig

+3/-2%, 1000 psig +3/-2%, 1010 psig +31-2%, 1025 psig +3/-2%, and 1035 psig +3/

2%.(1)

a. With less than four of the five MSSVs associated with each steam generator OPERABLE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(4) Two power-operated relief valves (PORVs) shall be operable OPERABLE during heatups and cooldowns when the RCS temperature is less than 515 0 F, and in Modes MODES 4 and 5 whenever the head is on the reactor vessel and the RCS is not vented through a 0.94 square inch or larger vent, to prevent violation of the pressure temperature limits designated by Figure 2-1.

a. With one PORV inoperable during heatups and cooldowns when the RCS temperature is less than 515 0 F, restore the inoperable PORV to operable OPERABLE within 7 days or be in cold shutdown COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b. With both PORVs inoperable during heatups and cooldowns when the RCS temperature is less than 515'F, be in cold shutdown COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15 Amendment No. 39,*7,54*1-46,161,89,207 TECHNICAL SPECIFICATIONS

c. With one PORV inoperable in-Modes MODES 4 or 5, within one hour ensure the pressurizer steam space is greater than 53% volume (50.6% or less actual level) and restore the inoperable PORV to operable OPERABLE within 7 days. If adequate steam space cannot be established within one hour, then restore the inoperable PORV to opefable OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the PORV cannot be restored in the required time, depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15a Amendment No. 39-,454,146*161,,1,89,207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

Basis The highest reactor coolant systemn pressure reached in any of the accidents analyzed resulted from a complete loss of turbine generator lcad without simultaneous reactor trip while operating at 1500 Mt~Ti rsucw3lc hntc25 zacft ii n the AS-lIE SectionlII upset pressure limit of 10% greater than the design pressure3.-The reactor is assumed to trip on a "I ligh Pressurizer Pressure" trip signal.

The pressurizer safety valves a requir Ie. to be calibrated to within v 1% of the specified setpoint value using A'ME Section Xl test methods. AS"MrE Section X" requires that valves in steam service use steam as the test medium, for establishing the setpoint. With the presence of a water-filled loop seal, establishing the valve setpoint with steam may result in in-situ valve actuation at pressures outside the 1 1%tolerance specified. Under transien conditions, it is expected that the valve(s) will actuate at no less than 4% below, nor gre"tet than 6% above, the specified setpoint, which is within the tolerance assumned in the safety analysis~tl Thesc analyses arc based on a mninimumn of any four of the five main stcamn safety valves on each main steamn header being OPERABLE.

The purpose of the two spring-loaded Pressurizer Safety Valves (PSV's) is to provide Reactor Coolant System (RCS) overpressure protection and thereby ensure that the Safety Limit for RCS pressure (i.e., 2750 psia) is not exceeded for analyzed accidents. The maximum RCS pressure transient for an analyzed accident is associated with a Loss of Load event( 2 ).

The TS 2.1.6(1) lift settings are determined during Surveillance Testing in accordance with ASME Code test methods. The ASME Code requires that valves in steam service use steam as the test medium for establishing the setpoint. The +1%I-3% tolerance range specified in TS 2.1.6(1) applies to opening pressures determined during Surveillance Testing. When the valves are installed in the system, the presence of a water-filled loop seal at the valve inlets may result in in-situ actuation at a pressure that differs from the actuation pressure with steam at the inlet. Comparative testing and analysis indicates that with a loop seal present, the opening pressure of these valves may be up to 1% lower than the opening pressure under normal test conditions. Opening pressures below the specified setpoints are not a concern with respect to the safety limit for RCS pressure. Analysis of a loss of load case involving elevated PSV opening pressures indicated that RCS pressures remained below the 2750 psia Safety Limit with PSV opening pressures up to 6% above the nominal setpoints. The valves are set to a tolerance of +/-1 % of setpoint using ASME Code test methods before being returned to service after testing. This allows for some setpoint variance over the surveillance interval.

The power-operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

2-15b Amendment No.

TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference

1. Control Element Drop times of all full-length CEA's Prior to reactor criticality after each 7.5.3 Assemblies removal of the reactor vessel closure head

2. Control Element Partial movement of all CEA's Q 7 Assemblies (Minimum of 6 in)

3. Pressurizer Safety Set-Point Verify each pressurizer R 7 Valves safety valve is OPERABLE in accordance with the Inservice Testing Program. Following testing, lift settings shall be 2485 psig

+/-1 % and 2530 psig +/-1% respectively.

4. Main Steam Safety Set Point R 4 Valves

5. DELETED

6. DELETED

7. DELETED

8. Reactor Coolant Evaluate D* 4 System Leakage

9. Diesel Fuel Supply Fuel Inventory D 8.4 10a. Charcoal and HEPA 1. In-Place Testing** 9.10 Filters for Control Charcoal adsorbers and HEPA On a refueling frequency or every 720 Room filter banks shall be leak hours of system operation or after each tested and show >99.95% complete or partial replacement of the Freon (R-11 or R-1 12) and charcoal adsorber/HEPA filter banks, or cold DOP particulates after any major structural maintenance on removal, respectively. the system housing or following significant painting, fire or chemical releases in a ventilation zone communicating with the system.

• Whenever the system is at or above operating temperature and pressure.

• • Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3-20 Amendment No. 5,24Q,81*860A,66,169,171

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued)

b. Annual Occupational Exposure Report. An annual occupational exposure report shall be submitted on or before April 30 of each year. The report shall consist of a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functionsA' e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling outages. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

c. Monthly Operating Report. Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Document Control Desk, with a copy to the appropriate Regional Office, no later than the fifteenth of each month following the calendar month covered by the report.

This monthly report shall also include a statement regarding any challenges or failures to the pr essurizerý power operated relief valves or safety valves occurring during the subjet 5.9.2 Reportable Event A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear Regulatory Commission for any event meeting the requirements of 10 CFR Part 50.73.

5.9.3 Special Reports Special reports shall be submitted to the appropriate NRC Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification where appropriate:

a. In-service inspection report, reference 3.3.

b. Tendon surveillance, reference 3.5.

c. Containment structural tests, reference 3.5.

d. DELETED

e. DELETED

f. DELETED

g. Materials radiation surveillance specimens reports, reference 3.3.

h. DELETED

i. Post-accident monitoring instrumentation, reference 2.21

j. Electrical systems, reference 2.7(2).

3/ This tabulation supplements the requirements of § 20.2206 of 10 CFR Part 20.

5-7 Amendment No. 9'24,"35,846** 5,86,89***,

++E),-3,*9,7 ++,47,-52,+EAqO~4ee1-6, 202 LIC-02-0130 Page 1 ATTACHMENT 3 Clean Copy Technical Specification 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves Applicability Applies to the status of the pressurizer and main steam safety valves.

Objective To specify minimum requirements pertaining to the pressurizer and main steam safety valves.

Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:

(1) Two pressurizer safety valves shall be OPERABLE in MODES 1 and 2, with lift settings of 2485 psig +1%/-3% and 2530 psig +t1%/-3% respectively.

(2) Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one OPERABLE safety valve shall be installed on the pressurizer.

However, when in at least the COLD SHUTDOWN condition, safety valve nozzles may be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification.

(3) At least four of the five Main Steam Safety Valves (MSSVs) associated with each steam generator shall be OPERABLE in MODES 1 and 2. Lift settings shall be at 985 psig +3/-2%, 1000 psig +31-2%, 1010 psig +3/-2%, 1025 psig +3/-2%, and 1035 psig

+3/-2%.(')

a. With less than four of the five MSSVs associated with each steam generator OPERABLE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(4) Two power-operated relief valves (PORVs) shall be OPERABLE during heatups and cooldowns when the RCS temperature is less than 515'F, and in MODES 4 and 5 whenever the head is on the reactor vessel and the RCS is not vented through a 0.94 square inch or larger vent, to prevent violation of the pressure-temperature limits designated by Figure 2-1.

a. With one PORV inoperable during heatups and cooldowns when the RCS temperature is less than 515 0 F, restore the inoperable PORV to OPERABLE within 7 days or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b. With both PORVs inoperable during heatups and cooldowns when the RCS temperature is less than 515'F, be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15 Amendment No. 39,47,54,146,1-6,1,189,207

c. With one PORV inoperable in MODES 4 or 5, within one hour ensure the pressurizer steam space is greater than 53% volume (50.6% or less actual level) and restore the inoperable PORV to OPERABLE within 7 days. If adequate steam space cannot be established within one hour, then restore the inoperable PORV to OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the PORV cannot be restored in the required time, depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15a Amendment No. 39,47,544,46,16*189,207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

Basis The purpose of the two spring-loaded Pressurizer Safety Valves (PSV's) is to provide Reactor Coolant System (RCS) overpressure protection and thereby ensure that the Safety Limit for RCS pressure (i.e., 2750 psia) is not exceeded for analyzed accidents. The maximum RCS pressure transient for an analyzed accident is associated with a Loss of Load event(2).

The TS 2.1.6(1) lift settings are determined during Surveillance Testing in accordance with ASME Code test methods. The ASME Code requires that valves in steam service use steam as the test medium for establishing the setpoint. The +1 %/-3% tolerance range specified in TS 2.1.6(1) applies to opening pressures determined during Surveillance Testing. When the valves are installed in the system, the presence of a water-filled loop seal at the valve inlets may result in in-situ actuation at a pressure that differs from the actuation pressure with steam at the inlet. Comparative testing and analysis indicates that with a loop seal present, the opening pressure of these valves may be up to 1% lower than the opening pressure under normal test conditions. Opening pressures below the specified setpoints are not a concern with respect to the safety limit for RCS pressure. Analysis of a loss of load case involving elevated PSV opening pressures indicated that RCS pressures remained below the 2750 psia Safety Limit with PSV opening pressures up to 6% above the nominal setpoints. The valves are set to a tolerance of +/-1% of setpoint using ASME Code test methods before being returned to service after testing. This allows for some setpoint variance over the surveillance interval.

The power-operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

2-15b Amendment No.

TABLE 3-5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS USAR Section Test Frequency Reference

1. Control Element Drop times of all full-length CEA's Prior to reactor criticality after each 7.5.3 Assemblies removal of the reactor vessel closure head

2. Control Element Partial movement of all CEA's Q 7 Assemblies (Minimum of 6 in)

3. Pressurizer Safety Verify each pressurizer safety valve R 7 Valves is OPERABLE in accordance with the Inservice Testing Program.

Following testing, lift settings shall be 2485 psig +/-1% and 2530 psig +/-1% respectively.

4. Main Steam Safety Set Point R 4 Valves

5. DELETED

6. DELETED

7. DELETED

8. Reactor Coolant Evaluate D* 4 System Leakage

9. Diesel Fuel Supply Fuel Inventory D 8.4 10a. Charcoal and HEPA 1. In-Place Testinq** 9.10 Filters for Control Charcoal adsorbers and HEPA On a refueling frequency or every 720 Room filter banks shall be leak hours of system operation or after each tested and show >99.95% complete or partial replacement of the Freon (R-11 or R-112) and charcoal adsorber/HEPA filter banks, or cold DOP particulates after any major structural maintenance on removal, respectively, the system housing or following significant painting, fire or chemical releases in a ventilation zone communicating with the system.

Whenever the system is at or above operating temperature and pressure.

• • Tests shall be performed in accordance with applicable section(s) of ANSI N510-1980.

3-20 Amendment No. 15,24,I28,1-6,166,169,171

5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued)

b. Annual Occupational Exposure Report. An annual occupational exposure report shall be submitted on or before April 30 of each year. The report shall consist of a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functionsY e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling outages. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

c. Monthly Operating Report. Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Document Control Desk, with a copy to the appropriate Regional Office, no later than the fifteenth of each month following the calendar month covered by the report.

5.9.2 Reportable Event A Licensee Event Report (LER) shall be submitted to the U.S. Nuclear Regulatory Commission for any event meeting the requirements of 10 CFR Part 50.73.

5.9.3 Special Reports Special reports shall be submitted to the appropriate NRC Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification where appropriate:

a. In-service inspection report, reference 3.3.

b. Tendon surveillance, reference 3.5.

c. Containment structural tests, reference 3.5.

d. DELETED

e. DELETED

f. DELETED

g. Materials radiation surveillance specimens reports, reference 3.3.

h. DELETED

i. Post-accident monitoring instrumentation, reference 2.21

j. Electrical systems, reference 2.7(2).

3/ This tabulation supplements the requirements of § 20.2206 of 10 CFR Part 20.

5-7 Amendment No. 9,24,35,28,46,75,86,8,9 Eq*,,++,,33,e41q-52,-+7,4e-+e4,,+,

7 202