LIC-02-0103, License Amendment Request, Relocation of Prestressed Containment Tendon Surveillances to the USAR

License Amendment Request, Relocation of Prestressed Containment Tendon Surveillances to the USAR

ML022900024
Person / Time
Site:	Fort Calhoun
Issue date:	10/08/2002
From:	Bannister D Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-02-0103, NUREG-1432, Rev 2
Download: ML022900024 (36)

Text

Dmmm Omaha Public Power Distnct 444 South 16th Street Mall Omaha NE 68102-2247 October 8, 2002 LIC-02-0103 U. S. Nuclear Regulatory Commission ATTN.: Document Control Desk Washington, DC 20555

References:

1. Docket No. 50-285

2. Improved Standard Technical Specification (ITS) for Combustion Engineering Plants, NUREG-1432, Revision 2

SUBJECT:

Fort Calhoun Station Unit No. 1 License Amendment Request, Relocation of Prestressed Containment Tendon Surveillances to the USAR Pursuant to 10 CFR 50.90, Omaha Public Power District (OPPD) hereby transmits an application for amendment to the Fort Calhoun Station Unit 1 (FCS) Operating License. Attachment 1 provides the No Significant Hazards Evaluation and the technical bases for this requested change to the Technical Specifications (TS). Attachments 2 and 3 contain marked-up and clean-typed Technical Specification pages reflecting the requested Technical Specification and Basis changes.

The proposed amendment relocates the requirements of TS 3.5(5) for testing Prestressed Concrete Containment Tendons to the FCS Updated Safety Analysis Report (USAR). This proposed amendment also adds a TS requirement for a Containment Tendon Testing Program as TS 5.21 consistent with that presented in Section 5.5 of Reference 2.

OPPD requests approval of the proposed amendment by March 15, 2003. OPPD requests 120 days to implement this amendment. No commitments are made to the NRC in this letter.

I declare under penalty of perjury that the foregoing is true and correct. (Executed on October 8, 2002)

Employment with Equal Opportunity 4171

U. S. Nuclear Regulatory Commission LIC-02-0103 Page 2 If you have any questions or require additional information, please contact Dr. R. L. Jaworski at (402) 533-6833.

Sincerely, D. J. Bannister Manager Fort Calhoun Station DJB/TRB/trb Attachments:

1. Fort Calhoun Station's Evaluation

2. Markup of Technical Specification Pages

3. Clean-Typed Technical Specification Pages 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-02-0103 Page 1 Attachment 1 Fort Calhoun Station's Evaluation For Relocation of Prestressed Containment Tendon Surveillances to the USAR

2.0 DESCRIPTION

OF PROPOSED AMENDMENT

3.0 BACKGROUND

4.0 REGULATORY REQUIREMENTS AND GUIDANCE

6.0 REGULATORY ANALYSIS

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION (NSHC)

8.0 ENVIRONMENTAL CONSIDERATION

9.0 PRECEDENCE

10.0 REFERENCES

LIC-02-0103 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.

Omaha Public Power District (OPPD) proposes to relocate the requirements of TS 3.5(5) for testing Prestressed Concrete Containment Tendons to the FCS Updated Safety Analysis Report (USAR). This proposed amendment also adds a TS requirement for a Containment Tendon Testing Program as TS 5.21 consistent with that presented in Section 5.5 of Reference 10.1. This is acceptable since testing of containment tendons in accordance with ASME Boiler and Pressure Vessel Code,Section IX, Subsections IWE and IWL is specified in 10 CFR 50.55a. This change eliminates duplication of federal regulations. Therefore, this system does not meet the criteria set forth in 10 CFR 50.36(c)(2)(ii) for inclusion in the TS, and the requirements will be relocated to the USAR.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT The proposed change is to delete TS 3.5(5) and its associated Bases in their entirety and relocate this information to the USAR. This proposed amendment also adds a TS requirement for a Containment Tendon Testing Program as TS 5.21 consistent with that presented in Section 5.5 of Reference 10.1. This program maintains the reporting requirements of TS 3.5(5).

3.0 BACKGROUND

The proposed amendment relocates the requirements of TS 3.5(5) for testing Prestressed Concrete Containment Tendons to the USAR. This proposed amendment also adds a TS requirement for a Containment Tendon Testing Program as TS 5.21 consistent with that presented in Section 5.5 of Reference 10.1. This is acceptable since testing of containment tendons in accordance with ASME Boiler and Pressure Vessel Code,Section IX, Subsections IWE and IWL is specified in 10 CFR 50.55a. This change eliminates duplication of federal regulations and can be made without an impact on public health and safety. Therefore, this system does not meet the criteria set forth in 10 CFR 50.36(c)(2)(ii) for inclusion in the TS, and the requirements will be relocated to the USAR.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE FCS was licensed for construction prior to May 21, 1971, and at that time committed to the preliminary General Design Criteria (GDC). These preliminary design criteria are contained in the FCS USAR Appendix G.

This activity complies with FCS Design Criterion 10, "Containment," which is similar to 10 CFR 50 Appendix A GDC 16, "Containment design." FCS Design Criterion 10 states

LIC-02-0103 Page 3 that containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public.

This activity also complies with FCS Design Criterion 40, "Missile Protection," which is similar to 10 CFR 50 Appendix A GDC 4, "Environmental and dynamic effects design bases." FCS Design Criterion 40 states that protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures.

This activity also complies with FCS Design Criterion 49, "Containment Design Basis,"

which is similar to 10 CFR 50 Appendix A GDC 50, "Containment design basis." FCS Design Criterion 49 states that the containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems.

This activity also complies with FCS Design Criterion 50, "NDT Requirement for Containment Material," which is similar to 10 CFR 50 Appendix A GDC 51, "Fracture prevention of containment pressure boundary." FCS Design Criterion 50 states that principal load carrying components of ferritic materials exposed to the external environment shall be selected so that their temperature under normal operating and testing conditions are not less than 30'F above nil ductility transition (NDT) temperature.

All of these FCS Design Criteria will continue to be satisfied after the change to relocate the requirements of TS 3.5(5) for testing Pre-Stressed Concrete Containment Tendons to the USAR.

5.0 TECHNICAL ANALYSIS

Evaluation The proposed amendment relocates the requirements of TS 3.5(5) for testing Prestressed Concrete Containment Tendons to the USAR. This proposed amendment also adds a TS requirement for a Containment Tendon Testing Program as TS 5.21 consistent with that presented in Section 5.5 of Reference 10.1. This is acceptable since testing of containment tendons in accordance with ASME Boiler and Pressure Vessel Code,Section IX, Subsections IWE and IWL is specified in 10 CFR 50.55a. This change eliminates duplication of federal regulations and can be made without an impact on public health and safety. Therefore, this system does not meet the criteria set forth in 10 CFR

LIC-02-0103 Page 4 50.36(c)(2)(ii) for inclusion in the TS, and the requirements will be relocated to the USAR.

The addition of a Containment Tendon Testing Program as TS 5.21 is considered administrative since it addresses those requirements in TS 3.5(5), which is consistent with Reference 10.1. The existing reporting requirements of TS 3.5(5) are maintained in the proposed TS 5.21.

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

The risk to the health and safety of the public as a result of relocating requirements of TS 3.5(5) for testing Prestressed Concrete Containment Tendons to the USAR is minimal.

6.0 REGULATORY ANALYSIS

The proposed amendment relocates the requirements of TS 3.5(5) for testing Prestressed Concrete Containment Tendons to the USAR. This proposed amendment also adds a TS requirement for a Containment Tendon Testing Program as TS 5.21 consistent with that presented in Section 5.5 of Reference 10.1. This is acceptable since testing of containment tendons in accordance with ASME Boiler and Pressure Vessel Code,Section IX, Subsections IWE and IWL is specified in 10 CFR 50.55a. This change eliminates duplication of federal regulations and can be made without an impact on public health and safety. Therefore, this system does not meet the criteria set forth in 10 CFR 50.36(c)(2)(ii) for inclusion in the TS, and the requirements will be relocated to the USAR. This complies with the regulatory requirements in FCS Design Criteria 10, 40, 49, and 50 by continuing to prevent damage to the containment structure.

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

Response: No

LIC-02-0103 Attachment I Page 5 The proposed change relocates requirements for testing Prestressed Concrete Containment Tendons that do not meet the criteria for inclusion in the TS set forth in 10 CFR 50.3 6(c)(2)(ii). The requirements for testing Prestressed Concrete Containment Tendons are being relocated from TS to the USAR, which will be maintained pursuant to 10 CFR 50.59, thereby reducing the level of regulatory control. The level of regulatory control has no impact on the probability or consequences of an accident previously evaluated. Therefore, the change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed change relocates requirements for testing Prestressed Concrete Containment Tendons that do not meet the criteria for inclusion in TS set forth in 10 CFR 50.36(c)(2)(ii). The change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The change will not impose different requirements, and adequate control of information will be maintained. This change will not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No The proposed change relocates requirements for testing Prestressed Concrete Containment Tendons that do not meet the criteria for inclusion in TS set forth in 10 CFR 50.36(c)(2)(ii). The change will not reduce a margin of safety since the location of a requirement has no impact on any safety analysis assumptions. In addition, the relocated requirements for testing Prestressed Concrete Containment Tendons remain the same as the existing TS. Since any future changes to these requirements or the surveillance procedures will be evaluated per the requirements of 10 CFR 50.59, there will be no reduction in a margin of safety.

Based on the above, 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 proposed amendment relocates the requirements of TS 3.5(5) for testing Prestressed Concrete Containment Tendons to the USAR. This proposed amendment also adds a TS

LIC-02-0103 Page 6 requirement for a Containment Tendon Testing Program as TS 5.21 consistent with that presented in Section 5.5 of Reference 10.1. This is acceptable since testing of containment tendons in accordance with ASME Boiler and Pressure Vessel Code,Section IX, Subsections IWE and IWL is specified in 10 CFR 50.55a. This change eliminates duplication of federal regulations. Therefore, this system does not meet the criteria set forth in 10 CFR 50.36(c)(2)(ii) for inclusion in the TS, and the requirements will be relocated to the USAR. The changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

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

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 off-site. Also, the TS change does not introduce any new effluents or significantly increase the quantities of existing effluents. As such, the change cannot significantly affect the types or amounts of any effluents that may be released off-site.

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 as a result of these 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 The proposed Technical Specifications are patterned after the Improved Standard Technical Specifications as described in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants" Reference 10.1. The NRC has approved specifications very similar to these proposed changes for the Palisades Nuclear Power Plant.

10.0 REFERENCES

10.1 NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants"

LIC-02-0103 Page 1 ATTACHMENT 2 Markup of Technical Specification Pages Bases Pages

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIRE1MENT*

3.5 Containment TeGsts (ontinued)

(5+ Sur'weillanee for Prestressing System

a. Sample SelectioGn The 210 dome tendons and 616 helc;al wall tendons shall be periodicall.

inspectcd for sympto~m~s of mnaterFial deterioration or prestressing force reduction.

inspec~tion hl e efre on four dome tendons, one from each layer an the cntrol dome tendon, and te helic-al wall tendons, five of each orientation inc~luding one conrol1 tendon ineach orientatkion.

The tendons to be inspected shall be randomly selected from the tendons which have not been tested in previous su11eillan1es, except for the cnroltendo Which shall be included in each surveillane sample seleGion to develop a historical trend in order to correlate the obser.'d data.

b. Visual llnsecti n The follow.ing visual inspections shall be performed:

(I) The cverinr surface of the containment shall be visually examined to detecat areas of large spa!!, severe scaling, D cracking in areas of 2 square feet oFr more, grease leakage, and other significant strutural d-et~erio-ratfion or disintegration.

(ii) For each sur.'eillance tendon, selecsted in accordance with 3.5(5)a., the tendon anchorage assembly hardware shall be visually inspected for signs of abnormal m1aterial behav;ir or wear.

(iii) The concrete SUrrounding the visually inspected tendon anchoragesshl be v.isually, inspected for signs of significant structural deterioration.

(i'1) The bottom grease cap, of all helical wall tendons shall be v.'isual! I inspected to detect grease leakage or grease cap deformations. Removal of grasecapsishot te necessar,' for thisinpco.

Pr~estress Monitoring Tests, Liftoff tests shall be performed on each tendon selected in accor~dance wfith 3.5(5)a. to monitor prestress. A~dditionally, the tests shall include the folloyA.'ig 3-45 Amendment No. 95,97,139,151,185

TECHNICAL SPECIFICATIONS "3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Teg-ts (Continued)

(i) Two helical wall tendons, one of each orientation, and one dome tendon, each randomly selected from their respective groups of sur'eillanc tendons, Ihall e*dletensionedan d for broken o damaged wires.

The control tendons shall NOT be included as tendons to be detensioned.

(i) During retcnsion~ing, simultaneous elongation and jackingfoc spaced levels of force between ze-r and the Ick Off force. The tWo intermediate stress levels shall be as near as practical to-the values shown on the initial stresn record for the respective tendon.

d. Tendon Material Tests and lnsvections One wire fromF each of two helical wall tendons, one of each orientation, andon deme tenpon, shall he removed Tor the following tesis and xmntos (i) EachF1 reved wire shall be examined over its entire length for any evence of corrosion r ot*her deteoration.

(i)Tensile tests rhall be made on at least three samnples of each wire, one cut f~rom each end and one cut ferom idlength. The samples shall bet maximum length pracatical for testing and the guage length for elongation shall be in accordance with ASTM E8 "Standard Test Methods for Tension Testing of Metallic Materials." The following infIIlation shall be obtained f each test 1rom (a) Yield Strength, (b) Ultimate tensile strength, and (c) Elongation at ultimate tensile strength 3-46 Amendment No. 95,97,119,139

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containmen~t Tpgts (Continued)

The tendons detensioned in accordance with 3.5(5)c.(i) mnay be the tendons from which the sample i*ra*eemoved. The cro tendons shall NOT be included as tendons to be dletensioned or have wires removed. lI addition, all vIIres found to be broken shall be emoved tensile testing and visual examination.

taken and analyzed according to the followin g*national standards:*

(i) To determine water content, ASTM D95, "Standard Test Methods for Water in Petroleum Products and Bituminous Materials by Distillation."

(ii) To determine reserve alkalinity, AST-M D974, "Standard Test Methodfo Acid and Base Number by Color Indicater Titration."

(iii) To determnine the concentration of w.ater soluble chlorides, AST-M D5142, "Standard Test Methods for Chloride Ion ;*i Water" (iv) To deteFmine the con.entration of w.ater solubl nitrates, ASTM D3867, "Standard Test Methods for Nitrito NItrate In WAIater."

(Y.) To determine the concentration of w.ater soluble sulfides, APHA 4500 ~

D. "Methylene Blue Method," Standard Metho-ds for EXWWaminain of Water adWaste Water, Seventeenth Editionr.

in addition to these tests, the amount of filler grease removed from and r-eplae into each surveillance tendon 6hall be recor-ded and wompared to assess grease leakage within the containment structure.

IF" 4L T. ,ce ,riteria

,-ance

()No i

evidence of significant strucatural deterioration of the conrGete inspected accordan.e with 3.5(.)b.(i) and 3..(5)b.(iii) which may affect the structural integrity of the containment structue can be detected.

3-47 Amendment No. 95,97,139,151,185

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued) detwerioatien which, when compared with past inspectionSsossrn Of eRden.c of an increase *r*uctural dleteroration which could affect the Containment's structural integrity. Evidene of oasmetic or superfical dleterioration, unless dletermnined by sound enginern Judgementtob significant, is not c~onsidered to bc significantstutua deteroain No evidence Of signifcant mnaterial degradatio ocrosion of tendo anchorage hardware can be detected.

if a.ny, g~re.'s, Ileakage ;s de.÷,-te~t l ,4..r9n visual

• exa-m;i-n'tio.n of* the ,-,-nta'inm rt exerAior surface, an investigation shall be made to dcteFmine the extcRt Gt potential reduction of Containment stuctuFral integity. An investigation shall also be made to determine which tendons could have lost the grease and whether the gre.ase lss has adversely affected their corrosio. protecton.

prestressing fOrce measured for each tendon liftoff tested in

(,.),The accr~rdance with 3.5(6)r.. shall be compared with the limits predicsted by USAR Fig 5.10 3. if the measured prestrcssing force of a selected tendoan is greater than the prescribed lower limit, the tendon is acceptable.

prescribed lower limit but greater than or equal to 95% of the prescriFbed lower limit, the tendon shall be tensioned to a prestress value greater than the prcscribed lower limit but less than 7-42 kips. .AfteF increasing the tendon's prestress the tendon will be considered acceptable.

3-48 Amendment No. 95,97,139,151,185

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests~g (Continucd) if the meaSUred preStressing for~e of a selected tendon is less than 95% oe the prescribed lower limit but greater than OF equal to 90% of the proscribcd locwer limit, two addiional tendons, one on each side of the first.

tendon, shall be liftoff tested. if the p.estressing foFres o.f eac-,h o-f the second and third tendons are greater than 95% of the prescribed !owet limit, all three tendons shall be tensioned to greater than the presrFibed lower limit, but less than 7-42 kips. After increasing the tendons' prestress, the tendoRns w;ill be c-onsidered acceptable. If the prestressing force oF either the second Or third tendons,is less than 95% of the prescribed !ower limit, liftoff tests shall be performned On additional tendons to determine the cause and extent Of such occurrence. This occurrencre shall be con sidered r~eportable per 3.5(5)g. if the measured prestressing force of a selected tendn is less than 90% of the prescribed low.er limit, the defect-ive tendo shall be fully inspected to dletermine the cseand extent Of such oGccurrence. This occurrence shall be considered reportable per 3.5(5)g.

if the average prestressing force- of nall mepaSUred tendons of a group (corrected for average condition) is found to be less than the presribe lower limit, an iesgaonshall be performed to determine the cause and extent of such an occurrence. Such an occrrUFence shall be considered reportable per 3.5(5)g.

if fromn consecutive surveillances the average mneasured prestressing force of a tendon group trends at a rate which would indicate that the loseo prestress would makethe average preStress of the group of tendons less than the prescI;rel lowr Ibed nexIsurweilanIe, additional lifto limitIbefore t-heI occurrence. Such an occurrence shall be considered reportable per 3.5(5)g.

(iii) if during the detensining and retenSioning of tendons in accordance with 3.5(5)G., the elongation corresponding to a specific load differs by moe than I 0%from that recorded during installation of the tendons, an investigation shall be mnade to ensure that the difference is not related to wire failures Or slippage of wires in anchorages. A difference Of MOre than 10% shall be considered reportable per 3.5(5)g.

3-49 Amendment No. 95,97,139,151,185

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment T-ests (Continued)

(iv) The minimnum accreptable ultimate tensile strength of the wire samples to be tenSile tested shall be 210,000 psi with a minimum elongation of 4% in accordance with ASTM .A4 21 65 for Type BA wire."rFailurwe inthe tensile test at strength or elongation values less than those specified shall be considered reportable per 3.5(5)9. Other conditions which indicate corrosion found by visual examination of the wire shall be considered repor-table per-3.5(6)g.

(Y) Results of the laboratory tests and examinations of the filler grease will be con~sider-ed acceptable if the following condcitions are met:

(a) i^.at*e* content - 10% by weight (b) Chlorides - 10 , ppFm (G) Nitrates -10 ppm (d) Sulfides 10 ppm (e) Reserwe alkalinity>

(Base numbers)

(DThe diffcencc~ between the amoeunt of grease injected into a tendon to replace the amounit whicvh IWAas removed during inspection shal not exceed 5%of the net tendon sheath (duct) volume when inec*edth*erigihl*installatio pressure.

(g) TheIlak of the prIesene of any free water.

"Thefai!ure to mneet any of the above cOncditions rf- the filler grease shall be considered reportable per 3.5(5)g.

g. Corrective Action and RevortinQ if the above acceptance criteria are not met, an immediate ivtgaonshall bee made to determine the cause(s) and extent Of the non confo~rmance to the criteria-,

and the results shall be reported-to the Commission within 90 days v:ia a speca report in accordance with Tech*i*al Specficatio 5.9..

3-50 Amendment No. 24,68,95,139,151,185

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Conta;inmnent Tests (Continued)

h. Test Frequenc','

The tendon prestressing system surweillancc shall be performed once ee~y5 eafrs Basis The containment is designed for an accident pressure of 60 psig.(2 ) While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a maximum temperature of about 1200 F.

With these initial conditions the temperature of the steam-air mixture at the peak accident pressure of 60 psig is 2880 F.

Prior to initial operation, the containment was strength-tested at 69 psig and then was leak tested. The design objective of the pre-operational leakage rate test has been established as 0.1% by weight for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. This leakage rate is consistent with the construction of the containment, which is equipped with independent leak-testable penetrations and contains channels over all inaccessible containment liner welds, which were independently leak-tested during construction.

Safety analyses have been performed on the basis of a leakage rate of 0.1% of the free volume per day of the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the maximum hypothetical accident. With this leakage rate, a reactor power level of 1500 MWt, and with minimum containment engineered safety systems for iodine removal in operation (one air cooling and filtering unit), the public exposure would be well below 10 CFR Part 100 values in the event of the maximum hypothetical accident. (3) The performance of an integrated leakage rate test and performance of local leak rate testing of individual penetrations at periodic intervals during plant life provides a current assessment of potential leakage from the containment.

The reduced pressure (5 psig) test on the PAL is a conservative method of testing and provides adequate indication of any potential containment leakage path. The test is conducted by pressurizing between two resilient seals on each door. The test pressure tends to unseat the resilient seals which is opposite to the accident pressure that tends to seat the resilient seals. A periodic test ensures the overall PAL integrity at 60 psig.

The integrated leakage rate test (Type A test) can only be performed during refueling shutdowns.

3-51 Amendment No. 68,97, 39,**5 *, 85

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (continued)

The frequency of periodic integrated leakage rate tests is based on several major considerations: (1) There is a low probability of leaks in the liner because of the test of leak-tightness of the welds during erection and conformance of the complete containment to a low leak rate at 60 psig during pre-operational testing, which is consistent with 0.1% leakage at design basis accident conditions and absence of any significant stresses in the liner during reactor operation. (2)

Periodic testing is conducted at full accident pressure, on those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves). A low value (0.60 La) of total leakage is specified as acceptable from penetrations and isolation valves. (3) The tendon stress surveillance program provides assurance that an important part of the structural integrity of the containment is maintained. (4) A review of leakage rates obtained during past containment integrated leakage rate testing is conducted to set appropriate frequency of performance. (5) Visual inspection of the containment structure is conducted every other refueling and prior to each Integrated Leakage Rate Test.

As left leakage prior to the first startup after performing a required leakage test is required to be < 0.6 La for combined Type B and C leakage, and < 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of < 1.0 La. At <

1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.

Integrity tests of the purge isolation valves are established to identify excessive degradation of the resilient seats of these valves. Simultaneous testing of redundant purge valves from a leak test connection accessible from outside containment provides adequate testing. The testing method is identical to the Type C purge isolation valve test performed in accordance with 10 CFR Part 50, Appendix J. For leakages found to be greater than 18,000 SCCM, repairs shall be initiated to ensure these valves meet the acceptance criteria.

A redu~tien in prsresn fem an Ghang copyi~lGnditions are exese for the prestressing system. Allowances have been made in the reactor building design fFr the rwedution and changes. Through Gmrrparsins between the documentd ins pcto4n results and the initial quality control reordsthe reducations in prestress and the physical changes, are trended to Verify excesse reductions Or changes do not occur or arc detected in a timely ma nner to be Gor~eoted.

3-52 Amendment No. 97,139485

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Contalinment Tests (Continued)

The prestressing system is a necessar,' strength element of the plant safegurd and it is desirable to confirmn that the allowances are not being exceeded. The technique chosen f*F surveillance is based on the rate Of change of prestressing force and physical conditions so that the surveillance can either confirmn that h allowances are SUfficient or require mnaintenanc~e befoe mnumlevels of pr~estressing force or physical conditions are reached. The end anhrg concr~ete is needed to maintain th rsrsgforces. The desg inetigations have concluded that the design is adequate and this has been confirmed by tests.

The prestesn seuence has shown that the end anchorage concrete can withstand loadslin.ex,.ncss oaf those which result when the tendons are anchored.

Fu~theF, the containment building w~.as pressure tested to 1.15 times the maximnum d§pesign peSS~

References (1) USAR, Section 5.9 (2) USAR, Section 5.1.1 (3) USAR, Section 14.15 3-53 Amendment No. 95,39

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.7 Safety Limit Violation 5.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The unit shall be placed in at least HOT SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b. The Safety Limit Violations shall be reported to the corporate officer responsible for overall plant nuclear safety and the Chairperson of the Safety Audit and Review Committee (SARC) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Plant Review Committee. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Chairperson of the Safety Audit and Review Committee and the corporate officer responsible for overall plant nuclear safety within 14 days of the violation.

5.8 Procedures 5.8.1 Written procedures and administrative policies shall be established, implemented and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978;

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Fire Protection Program implementation; and

d. All programs specified in Specification 5.11 through 5.4021.

5.8.2 Temporary changes to procedures of 5.8.1 above may be provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License.

5-5 Amendment No. 9,19,38,84,90, 115,149,157,160,184,-202

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 functions,- 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 totalling 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 pressurizer power operated relief valves or safety valves occurring during the subject month.

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 5.21.

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. 9,2 ,35,28,46,75,86,89,99, 440,4 43,4 4 ,433,4 47,4 52,4 57,4 60,464,485, 202

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.19 Containment Leakage Rate Testing Program (Continued)

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is _<1.0 La. During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are _*0.60 La Maximum Pathway Leakage Rate (MXPLR) for Type B and C tests and _50.75 La for Type A tests.

b. Personnel Air Lock testing acceptance criteria are:

(1) Overall Personnel Air Lock leakage is _0.1 La when tested at > Pa.

(2) For each PAL door, seal leakage rate is *0.01 La when pressurized to Ž5.0 psig.

c. Containment Purge Valve (PCV-742A/B/C/D) testing acceptance criterion is:

For each Containment Purge Valve, leakage rate is < 18.000 SCCM when tested at >__

Pa.

d. Ifat any time when containment integrity is required and the total Type B and C measured leakage rate exceeds 0.60 La Minimum Pathway Leakage Rate (MNLPR), repairs shall be initiated immediately. If repairs and retesting fail to demonstrate conformance to this acceptance criteria within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then containment shall be declared inoperable.

The provisions of Specification 3.0.1 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 3.0.4 are applicable to the Containment Leakage Rate Testing Program.

[5.20 Bases ControlProgram(proposedto be added by submittaldated July 23, 2002)]

5.21 Containment Tendon Testing Program This program provides controls for monitoring any tendon degradation in prestressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Containment Tendon Testing Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, 1989.

5-17 Amendment No. 185n,20

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.21 Containment Tendon Testing Program(Continued)

The provisions of TS 3.0.1 and TS 3.0.5 [(proposed to be added by submittal dated July 23, 2002)] are applicableto the Containment Tendon Testing Programinspection frequencies.

If the acceptance criteriaare not met, an immediate investigationshall be made to determine the cause(s) and extent of the non-conformance to the criteria,and the results shall be reported to the Commission within 90 days via a specialreportin accordance with Technical Specification 5.9.3.

5-18 Amendment No.

LIC-02-0103 Page 1 ATTACHMENT 3 Clean-Typed Technical Specification Pages Bases Pages

TECHNICAL SPECIFICATIONS NOT USED 3-45 Amendment No. 95,97,139,151,185

TECHNICAL SPECIFICATIONS NOT USED 3-46 Amendment No. 95,97,119,139

TECHNICAL SPECIFICATIONS NOT USED 3-47 Amendment No. 95,97,139,161,185

TECHNICAL SPECIFICATIONS NOT USED 3-48 Amendment No. 05,97,139,151,1,*

TECHNICAL SPECIFICATIONS NOT USED 3-49 Amendment No. 95,97,139,4 54,4 185

TECHNICAL SPECIFICATIONS NOT USED 3-50 Amendment No. 24,68,95,1 39,161,185

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS Basis The containment is designed for an accident pressure of 60 psig.(2) While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a maximum temperature of about 1200 F.

With these initial conditions the temperature of the steam-air mixture at the peak accident pressure of 60 psig is 2880 F.

Prior to initial operation, the containment was strength-tested at 69 psig and then was leak tested. The design objective of the pre-operational leakage rate test has been established as 0.1% by weight for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. This leakage rate is consistent with the construction of the containment, which is equipped with independent leak-testable penetrations and contains channels over all inaccessible containment liner welds, which were independently leak-tested during construction.

Safety analyses have been performed on the basis of a leakage rate of 0.1% of the free volume per day of the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the maximum hypothetical accident. With this leakage rate, a reactor power level of 1500 MWt, and with minimum containment engineered safety systems for iodine removal in operation (one air cooling and filtering unit), the public exposure would be well below 10 CFR Part 100 values in the event of the maximum hypothetical accident. (3) The performance of an integrated leakage rate test and performance of local leak rate testing of individual penetrations at periodic intervals during plant life provides a current assessment of potential leakage from the containment.

The reduced pressure (5 psig) test on the PAL is a conservative method of testing and provides adequate indication of any potential containment leakage path. The test is conducted by pressurizing between two resilient seals on each door. The test pressure tends to unseat the resilient seals which is opposite to the accident pressure that tends to seat the resilient seals. A periodic test ensures the overall PAL integrity at 60 psig.

The integrated leakage rate test (Type A test) can only be performed during refueling shutdowns.

3-51 Amendment No. 68,97,139,151,185

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (continued)

The frequency of periodic integrated leakage rate tests is based on several major considerations: (1) There is a low probability of leaks in the liner because of the test of leak-tightness of the welds during erection and conformance of the complete containment to a low leak rate at 60 psig during pre-operational testing, which is consistent with 0.1% leakage at design basis accident conditions and absence of any significant stresses in the liner during reactor operation. (2)

Periodic testing is conducted at full accident pressure, on those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves). A low value (0.60 La) of total leakage is specified as acceptable from penetrations and isolation valves. (3) The tendon stress surveillance program provides assurance that an important part of the structural integrity of the containment is maintained. (4) A review of leakage rates obtained during past containment integrated leakage rate testing is conducted to set appropriate frequency of performance. (5) Visual inspection of the containment structure is conducted every other refueling and prior to each Integrated Leakage Rate Test.

As left leakage prior to the first startup after performing a required leakage test is required to be < 0.6 La for combined Type B and C leakage, and < 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of < 1.0 La. At <

1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.

Integrity tests of the purge isolation valves are established to identify excessive degradation of the resilient seats of these valves. Simultaneous testing of redundant purge valves from a leak test connection accessible from outside containment provides adequate testing. The testing method is identical to the Type C purge isolation valve test performed in accordance with 10 CFR Part 50, Appendix J. For leakages found to be greater than 18,000 SCCM, repairs shall be initiated to ensure these valves meet the acceptance criteria.

References (1) USAR, Section 5.9 (2) USAR, Section 5.1.1 (3) USAR, Section 14.15 3-52 Amendment No. 97,439,85

TECHNICAL SPECIFICATIONS NOT USED 3-53 Amendment No. 95,139

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.7 Safety Limit Violation a.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The unit shall be placed in at least HOT SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b. The Safety Limit Violations shall be reported to the corporate officer responsible for overall plant nuclear safety and the Chairperson of the Safety Audit and Review Committee (SARC) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Plant Review Committee. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Chairperson of the Safety Audit and Review Committee and the corporate officer responsible for overall plant nuclear safety within 14 days of the violation.

5.8 Procedures 5.8.1 Written procedures and administrative policies shall be established, implemented and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978;

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Fire Protection Program implementation; and

d. All programs specified in Specification 5.11 through 5.21.

5.8.2 Temporary changes to procedures of 5.8.1 above may be provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License.

5-5 Amendment No. 9,49,38,84, 145,149,157,160,184, 202

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 functions, 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 totalling 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 pressurizer power operated relief valves or safety valves occurring during the subject month.

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 5.21.

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. 9,24,35,28,46,75,86,89,99, 110,143,149,133,447,452,457,160,164,485,202

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.19 Containment Leakage Rate Testing Program (Continued)

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is _<1.0 La. During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are _<0.60 La Maximum Pathway Leakage Rate (MXPLR) for Type B and C tests and *0.75 La for Type A tests.

b. Personnel Air Lock testing acceptance criteria are:

(1) Overall Personnel Air Lock leakage is *_0.1 La when tested at >_Pa.

(2) For each PAL door, seal leakage rate is *<0.01 La when pressurized to >_5.0 psig.

c. Containment Purge Valve (PCV-742A/B/C/D) testing acceptance criterion is:

For each Containment Purge Valve, leakage rate is < 18.000 SCCM when tested at >_ Pa.

e. If at any time when containment integrity is required and the total Type B and C measured leakage rate exceeds 0.60 La Minimum Pathway Leakage Rate (MNLPR), repairs shall be initiated immediately. If repairs and retesting fail to demonstrate conformance to this acceptance criteria within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then containment shall be declared inoperable.

The provisions of Specification 3.0.1 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 3.0.4 are applicable to the Containment Leakage Rate Testing Program.

f5.20 Bases Control Proqram (proposedto be added by submittal datedJuly 23, 2002)]

5.21 Containment Tendon Testing Program This program provides controls for monitoring any tendon degradation in prestressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Containment Tendon Testing Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3,1989.

5-17 Amendment No. 185 202

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.21 Containment Tendon Testing Program (Continued)

The provisions of TS 3.0.1 and TS 3.0.5 are applicable to the Containment Tendon Testing Program inspection frequencies.

If the acceptance criteria are not met, an immediate investigation shall be made to determine the cause(s) and extent of the non-conformance to the criteria, and the results shall be reported to the Commission within 90 days via a special report in accordance with Technical Specification 5.9.3.

5-18 Amendment No.