ML20003G191

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Tech Spec Table 4.1-3 for Min Sampling Frequency & Sections 4.4.2 for Containment Structural Integrity & 4.8 for MSIVs
ML20003G191
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 04/24/1981
From:
METROPOLITAN EDISON CO.
To:
Shared Package
ML20003G185 List:
References
NUDOCS 8104280449
Download: ML20003G191 (5)
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Text

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TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Item Check Frequency Reactor Coolant Radio-Chemical Anal determination (2)ysis (1) Monthly

1. a.
b. I Sesi-annually
c. 15 Min. Gross Degassed 5 times / week when Tavg Beta-Gamma Activity (1) is greater than 2000F
d. Tritium Radioactivity Monthly
e. Chemistry (C1, F and 0 3) 5 times / week when Tavg is greater than 2000F
f. Boron Concentration 2 times / week
2. Borsted Water Storage Boron Concentration Weekly and after each Tank Water Sample makeup when reactor coolant system pressure is greater than 300 psig orTavg is greater than 200 F
3. Core Flooding Tank Boron Concentration Monthly and after each Water Sample makeup when RCS pressure is greater than 700 psig
4. Spent Fuel Pool Boron Concentration Monthly and after each Water Sample makeup
5. Secondary Coolant a. 15 Min. Gross Degassed Weekly when reactor Beta-Gamma Activity coolant system pressure is greater than 300 psig or Tavg is greater than
b. Iodine Analysis (3)
6. Boric Acid Mix Tank Boron Concentration Twice weekly (4) l or Reclaimed Boric Acid Tank
10. Sodium Hydroxide Concentration Quarterly and after Tank each makeup
11. Sodium Thiosulphate Concentration Quarterly and after Tank each makeup
12. Condenser Partition 1131 Partition Factor Once if primary / secondary Factor leakage developes, i.e.:

Gross Beta-Gamma on secondary side of OTSG is greater than 2 x 10-8 microcuries per cc and evidence of fission products is present 8104280 %

4-9

,- .- .~- . - - -

(1) When radioactivity level is greater than 10 percent of the limits of Specification 3.1.4, the sampling frequency shall be increased to a minimum of 5 times per week.

(2) 5 determination will be started when the 15 minute gross degassed beta-gamma activity analysis indicates greater than 10 pCi/ml and will be redetermined each 10 pCi/mi increase in the 15 minute gross degassed beta-gamma activity analysis. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95 percent of radionuclides in reactor coolant with half lives of 2 30 minutes.

(3) When the 15 minute gross degassed activity increases by a factor of two above background, an iodine analysis will be made and performed thereafter when the 15 minute gross degassed beta-gamma activity increases by 10 percent.

(4) The surveillance of either the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank is not necessary when that respective tank is empty.

4-10

4.4.2 Structural Integrity Specification 4.4.2.1 Inservice Tendon Surveillance Requirements The surveillance program for structural integrity and corrosion pro-tection conforms to the recommendations of the U. S. Atomic Energy Commission Regulatory Guide 1.35, proposed Revision 3, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures." The detailed surveillance program for the prestressing system tendons shall be based on periodic inspection and mechanical tests to be performed on selected tendons, as specified hereafter.

4.4.2.1.1 Containment Tendons Tendon surveillance was completed for one, three and five years following initial structural integrity using a Tech. Spec. based on Regulatory Guide 1.35 Rev. 1. The containment tendon structural integrity shall be demonstrated at five year intervals thereaf ter by:

a. Determining that for a representative sample
  • of at least 23 tendons (6 dome, 7 vertical, and 10 hoop) each tendon has a lift off force equalling, or exceeding, its lovar limit predicted for the time of the test as defined in NRC Regulatory Guide 1.35,

" Inservice Inspection for Ungrouted Tendons in Prestressed Concrete Containments", Proposed Revision 3, April,1979.

If the lift off force of a selected tendon in a group lies between the prescribed lower limit and 90% of that limit, one tendon on each side of this tendon shall be checked for their lift off forces.

If the lif t off forces of the adjacent tendons are equal to, or greater than, their prescribed lower litiis at the time of the test, the single deficiency shall be considered unique and acceptable. If the lif t off force of either adjacent tendons lies below the prescribed lower limit for that tendon, the condition is report-able per T.S. 6.9.2.A3.

If the lift off force of any one tendon lies belov 90% of its pre-scribed lower limit, the tendon shall be considered a defective tendon. It shall be completely detensioned and a determination made as to the cause of the occurrence. The condition is reportable per T.S. 6.9.2,A3.

If the inspections performed at one, three, and five years indicate no abnormal degradation of the post-tensioning system, the number of tendons checked for lift off force during subsequent tests may be reduced to a representative sample of at least 11 tendons (3 dome,

, 3 vertical, and 5 hoop).

l F

  • For each inspection, the tendons shall be selected on a random but representative basis so that the sample group will change somewhat for each inspection; however, l to develop a history of tendon performance and to correlate the observed data one l tendon from each group (dome, vertical, and hoop) may be kept unchanged af ter the j initial selection.

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4-35 l

l

4.8 MAIN STEAM IS01.ATION VALVES Applicability-Applies to the periodic testing of the main steam isolation valves.

Objective To specify the minimum frequency and type of tests to be applied to the main steam isolation valves.

Specification 4.8.1 A check of valve stam movement, up to 10 percent, shall be performed on a monthly basis when the unit is operational and under normal flow and load conditions.

4.8.2 The matt steam isolation valves shall be tested at intervals not to exceed the normal refueling outage. Closure time of

< 120 seconds shall be verified. This test will be performed

[

under no flow and no load conditions.

Bases Since a portion of the main steam lines and the steam lines to the main feed pump turbines are located in the turbine hall which is not protected against hypothetical tornado, missile, or aircraf t incident; main steam isolation stop check valves are provided and located in the hardened portion of the interme.diate building. These stop check valves are remotely closed by the operator frog the control room, close in less than two minutes, and are tight closing til for long term ccatainment isolation.

Their ability to close upon signal should be verified at intervals not to exceed each scheduled refueling shutdown, and valve stem freedom should be checked on a monthly basis.

References l (1) FSAR, Section 10.2.1.3 i

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4-54 L

Metropolitan Edison Company g Post Office Box 480 II - A . Middletown, Pennsylvania 17057 Writer's oirect Dial Number April 24, 1981 LlL 081 Office of Nuclear Reactor Regulation Attn: R. W. Reid, Chief Operating Reactors Branch No. 4 U. S. Nuclear Regulatory Coc=tission Washington, D.C. 20555

Dear Sir:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 99 Enclosed are three signed originals (sixty conformed copies sent separately) of Technical Specification Change Request No. 99 requesting amend ent to Appendix A of Operating License No. DPR-50.

Also enclosed is one signed copy of Certificate of Service for proposed Technical Specification Change Request No. 99 to the chief executives of the township and county in which the facility is located.

Sincerely,

h. D. ukill Director, Dil-1 RCA:WSS:Ima

Enclosures:

1) Technical Specification Change Regt;est No. 99
2) Certificate of Service for Technical Specification Change Request No. 99
3) Check No.-Check will be forwarded under separate cover Metreocuan Ecison Cemcany is a Memter of the General PucLc Ltat et System

METROPOLITAN EDISON COMPAh7 JERSEY CENTRAL PO'a*ER & LIGHT COMPAhT AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 99 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No.

DPR-50 for Three Mile Island Nuclear Station, Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

METROPOLITAN EDISON COT.Ahi

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By Direhtor, Dil-1 H. D. Hukill l Sworn and subscribed to me this O V day of A , 1981.

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i UNITED STATES OF AMERICA NUCLEAR REGULATORY CO) MISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 99 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Commission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:

Mr. Donald Hoover, Chairman Mr. John E. Minnich, Chair: nan Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County ,

R.D. #1, Geyers Church Road Dauphin County Courthouse l

Middletown, PA 17057 Harrisburg, PA 17120 i

I METROPOLITAN EDISON COMPANY I

l By Dirdctor, TMI-l H. D. Hakill Dated: April 24, 1981 4

Three Mile Island Nuclear Station, Unit 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 99 The Licensee requests that the attached pages 111, 4-9, 4-10, 4-35, 4-51, and 4-54 replace the respective existing Tech. Spec. pages.

Reason for Change Request These administrative changes are requested for the following reasons:

99.1 - Specification 4.11 " Site Environmental Radioactivity Survey" refers the reader to Section 6.4 of Appendix B. There is no such section in Appendix B. Therefore, this section should be deleted.

99.2 Specification 4.8.1 states that a closure time of approximately 112 sec. shall be verified. The TMI-l FSAR Section 10.2.1.3 states < 120 sec. is the limit. The erroneous time of 112 sec.

I was obtained from a test procedure, which was later corrected to 120 sec. Therefore, the closure time in the Technical Speci-fication should be revised to < 120 sec.

  • 99.3 - Table 4.1-3 item 6 requires a twice per week baron concentration check of the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank.

A footnote should be added to this specification to allow relief from this sampling when the tanks are empty.

99.4 - Specification 4.4.2.1.1 " Containment Tendons" states that only the tendon surveillance done at one and three years following initial structural integrity used Regulatory Guide 1.35 Rev.1. This statement was correct when the Change Request was submitted for NRC review. However, the five-year surveillance was performed prior to receipt of Amendment 59 and was therefore also done per Regulatory Guide 1.35 Rev. 1. Therefore, the above referenced section has been rewritten to reflect what actually occurred.

Safety Analysis Justifying Change Because these changes are administrative in nature, they have no effect on nuclear safety. Therefore, no safety analysis is necessary.

License Amendment Classification (10CFR 170.22)

These changes are administrative in nature. Therefore, enclosed please find the proper remittance of $1200.00.

TABLE OF CONTENTS Section h 4.4.2 STRUCTURAL INTEGRITY 4-35

.4.4.3 HYDROGEN PURGE SYSTEM 4-37 4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER. EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4-39 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 i 4.5.2 EMERGENCY CORE COOLING SYST E 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEAT REMOVAL SYSTEM LEARAGE 4-45 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS 4-48 l 4.7.2 CONTROL ROD PROGRAM VERIFICATION 4-50 4.8 MAIN STEAM ISOLATION VALVES 4-51 4.9 EMERGENCY FEEDWATER PUMPS PERIODIC TESTING 4-52 4.9.1 TEST 4-52 4.9.2 ACCEPTANCE CRITERIA 4-52 4.10 REACTIVITY ANOMALIES 4-53 4.11 INTENTIONALLY BLANK 4-54 l 4.12 AIR TREATMENT SYSTEMS 4-55 i 4.12.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM 4-55 4.12.2 REACTOR BUILDING PURGE AIR TREAIMENT SYSTEM 4-55b 4.12.3 AUXILIARY AND FUEL HANDLING EXHAUST AIR TREATMENT SYSTEM 4-55d.

4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 REACTOR BUILDING PURGE EXHAUST SYSTEM 4-57 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 l 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59

, 4.17 SHOCK SUPPRESSORS (SNUBBERS) 4-60 4.18 FIRE PROTECTION SYST ES 4-72 4.18.1 FIRE PROTECTION INSTRUMENTS 4-72 4.18.2 FIRE SUPPRESSION WATER SYSTEM 4-73 4.18.3 DELUGE / SPRINKLER SYSTEM 4-74 4.18.4 CO2 SYSTEM 4-74

4.18.5 HALON SYSTEMS 4-75 l 4.18.6 HOSE STATIONS 4-76 l 4.19 OTSG TUBE INSERVICE INSPECIION 4-77 i 4.19.1 STEAM GENERATOR SAMPLE SELECTION l AND INSPECTION METHODS 4-77 l 4.19.2 STEAM GENERATOR TUBE SAMPLE SELECTION I

AND INSPECIION 4-77 4.19.3 INSPECTION FREQUENCIES 4-79 4.19.4 ACCEPTANCE CRITERIA 4-80 l

4.19.5 REPORTS 4-81 4.20 REACTOR BUILDING AIR TREATMENT 4-86 i 5 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 ,,

5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4

- 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTARE TUNNEL FIRE PROTECTION SYSTEMS 5-8 111

_ . - . . _ _ _ _ _ . , . . _ . . _ . . ~

TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Item Check Frequency

1. Reactor Coolant a.
b. E determination (2)ysis (1) Monthly Radio-Chemical Anal Semi-annually
c. 15 Min. Gross Degassed 5 times / week when Tavg Beta-Gamma Activity (1) is greater than 2000F
d. Tritium Radioactivity Monthly
e. Chemistry (C1, F and 0 2) 5 times / week when Tavg is greater than 2000F
f. Boron Concentration 2 times / week
2. Borated Water Storage Boron Concentration Weekly and after each Tank Water Sample makeup when reactor coolant system pressure is greater than 300 psig or Tavg is greater than 200 F
3. Core Flooding Tank Boron Concentration Monthly and after each Water Sample makeup when RCS pressure is greater than 700 psig
4. Spent Fuel Pool Boron Concentration Monthly and after each

, Water Sample makeup

5. Secondary Coolant a. 15 Min. Gross Degassed Weekly when reactor Beta-Gamma Activity coolant system pressure is greater than 300 psig or Tavg is greater than 2000F
b. Iodine Analysis (3)
6. BorAc Acid Mix Tank '

Boron Concentration Twica weekly (4) l

[ or Reclaimed Boric Acid Tank

10. Sodium Hydroxide Concentration Quarterly and after Tank each makeup i

l 11. Sodium Thiosulphate Concentration Quarterly and after l Tank each makeup

12. Condenser Partition Il31 Partition Factor once if primary / secondary Factor leakage developes, i.e.:

Gross Beta-Gamma on secondary side of OTSG is greater than 2 x 10-8 microcuries per cc and evidence of fission products is present 4-9

(1) When radioactivity level is greater than 10 percent of the limits of Specification 3.1.4, the sampling frequency shall be increased to a minimum of 5 times per week.

(2) E determination will be started when the 15 minute gross degassed beta-gamma activity analysis indicates greater than 10 pCi/mi and will be redetermined each 10 uCi/mi increase in the 15 minute gross degassed beta-gamma activity analysis. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95 percent of radionuclides in reactor coolant with half lives of 2 30 minutes.

(3) When the 15 minute gross degassed activity increases by a factor of two above background, an iodine analysis will be made and perfo-wd thereafter when the 15 minute gross degassed beta-gamna activity increases by 10 percent.

(4) The surveillance of either the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank is not necessary whet. that respective tank is empty.

i 4

10

4.4.2 Structural Integrity Specification 4.4.2.1 Inservice Tendon Surveillance Requirements The surveillance program for structural integrity and corrosion pro-tection conforms to the recommendations of the U. S. Atomic Energy Commission Regulatory Guide 1.35, proposed Revision 3, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures." The detailed surveillance program for the prestressing syst2m tendons shall be based on periodic inspection and mechanical tests to be performed on selected tendons, as specified hereafter.

4.4.2.1.1 Containment Tendons Tendon surveillance was completed for one, three and five years following initial structural integrity using a Tech. Spec. based on Regulatory Guide 1.35 Rev.1. The containment tendon structural it:egrity shall be demonstrated at five year intervals thereafter by:

a. Determining that for a representative sample
  • of at least 23 tendons (6 dome, 7 vertical, and 10 hoop) each tendon has a lif t off force equalling, or exceeding, its lower limit praJicted for the time of the test as defined in NRC Regulatory Guide 1.35,

" Inservice Inspection for Ungrouted Tendons in Prestressed Concrete Containments", Proposed Revision 3, April,1979.

If the lift off force of a selected tendon in a group lies between the prescribed lower limit and 90% of that limit, one tendon on each side of this tendon shall be checked for their lift off forces.

If the lif t off forces of the adjacent tendons are equal to, or greater than, their prescribed lower limits at the time of the test, the single deficiency shall be considered unique and acceptable. If the lif t off force of either adjacent tendons lies below the prescribed lower limit for that tendon, the condition is report-able per T.S. 6.9.2.A3.

If the lift off force of any one tendon lies below 90% of its pre-scribed lower limit, the tenden shall be considered a defective tenden. It shall be completely detensioned and a determination made as to the cause of the occurrence. The condition is reportab?e per T.S. 6.9.2.A3.

! If the inspections performed at cue, three, and fita years indicate no abnormal degradation of the post-tensioning system, the number of tendons checked for lift off force during subsequen: tests may be reduced to a representative sample of at least 11 tendons (3 dome, 3 vertical, and 5 hoop).

  • For each inspection, the tendons shall be selected on a random but representative basis so that the sample group will change somewhat for each inspection; however, to develop a history of tendon performance and to correlate the observed data one tendon from each group (dome, vertical, and hoop) may be kept unchanged after the initial selection.

4-35 1

4.8 MAIN STEAM ISOLATION VALVES Applicability Applies to the periodic testing of the main steam isolation valves.

~

] Objective To specify the minimn= frequency and type of tests to be applied to the main steam isolation valves.

Specification 4.8.1 A check of valve stem movement, up to 10 percent, shall be performed on a monthly basis when the unit is operational and under normal flow and load conditions.

4.8.2 The main steam isolation valves shall be tested at intervals not to exceed the normal refueling outage. Closure time of

< 120 seconds shall be verified. This test will be performed

[

under no flow and no load conditions.

7 Bases l Since a portion of the main steam lines and the steam lines to the main feed pump turbines are located in the turbine hall which is not protected against hypothetical tornado, missile, or aircraft incident; main steam isolation stop check valves are provided and located in the hardened portion of the intermediate building. These stop check valves are remotelyclosedbytheoperatorfroyphecontrolroon,closeinlessthan two minutes, and are tight closing for long term containment isolation.

Their ability to close upon signal-should be verified at intervals not to exceed each scheduled refueling shutdown, and valve stem freedom should be checked on a monthly basis.

References '

(1) FSAR, Section 10.2.1.3

+

d 4-51

0 &

INTENTIONALLY BLANK 4-54

Metropolitan Edison Company Post Office Box 480 il N ' Middletown, Pennsylvania 17057 Wnter's Direct Diat Number April 24, 1981 LlL 081 Office of Nuclear Reactor Ragulation Attn: R. W. Reid, Chief Operating Raaccors Branch No. 4 U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Sir:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-Su Docket No. 50-289 Technical Specification Change Request No. 99 Enclosed are three signed originals (sixty confor=ed copies sent separately) of Technical Specification Change Request No.99 requesting amendment to Appendix A of Operating License No. DPR-50.

Also enclosed is one signed copy of Certificate of Service for proposed Technical Specification Change Request No.99 to the chief executives of the township and county in which the facility is located.

Sincerely, l

. D. kill Director, TM1-1 l

RCA:W3S:1ma

Enclosures:

1) Technical Specification Change Request No. 99
2) Certificate of Service for Technical Specification change Request No. 99
3) Check No.-Check will be forwarded under separate cover MetroccLtan Ecscn Correany :s a Memcer et tre General Puct Ut l t.es Spem

METROPOLITAN EDISON COMPANY JERSEY CEN RAL POk'ER & LIGHT COMPAhT AND PENNSYLVANIA ELECTRIC COMPANY I"dREE MILE ISLLND NUCLEAR STATION, UNIT 1 Operating License No. DPR-50 Docke. No. 50-289 Technical Specification Change Request No. 99 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No.

DPR-50 for Three Mile Island Nuclear Station, Unit 1. As a part of this request, proposed replacement pages for Appendix A are also l included.

I i

METROPOLITAN EDISON COMPAhi

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i ' Director, IMI-l j H. D. Hukill Swor and subscribed to :ne this ,W day of N/1 , 1981.

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 99 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with I

the U. S. Nuclear Regulatory Commission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:

Mr. Donald Hoover, Chairman Mr. John E. Minnich, Chair: nan Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R.D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 METROPOLITAN EDISON COMPANY l

By

~

' Dirdctor, TMI-l H. D. Hukill Dated: April 24, 1981

Three Mile Island Nuclear Station, Unit 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 99 The Licensee requests that the attached pages iii, 4-9, 4-10, 4-35, 4-51, and 4-54 replace the respective existing Tech. Spec. pages.

Reason for Change Request These administrative changes are requested for the following reasons:

99.1 - Specification 4.11 " Site Environmental Radioactivity Survey" refers the reader to Section 6.4 of Appendix B. There is no such section in Appendix B. Therefore, this section should be deleted.

99.2 Specification 4.8.1 states that a closure time of approximately 112 sec. shall be verified. The TMI-l FSAR Section 10.2.1.3 states < 120 sec. is the limit. The erroneous time of 112 sec.

was obtained fr6a a test procedure, which was later corrected to 120 sec. Therefore, the closure time in the Technical Speci-fication should be revised to < 120 sec.

99.3 - Table 4.1-3 item 6 requires a twice per week boron concentration check of the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank.

A footnote should be added to this specification to allow relief from this sampling when the tanks are empty.

99.4 - Specification 4.4.2.1.1 " Containment Tendens" states that only the tendon surveillance done at one and three years following initial structural integrity used Regulatory Guide 1.35 Rev.1. This statement was correct when the Change Request was submitted for NRC review. However, the five-year surveillance was performed prior to receipt of Amendment 59 and was therefore also done per Regulatory Guide 1.35 Rev. 1. Therefore, the above referenced section has been rewritten to reflect what actually occurred.

Safety Analysis Justifying Change s

Because these changes are administrative in nature, they have no effect on nuclear safety. Therefore, no safety analysis is necessary.

License Amendment Classification (10CFR 170.22)

These changes are administrative in nature. Therefore, enclosed please find the proper remittance of $1200.00.

e

+ - ,n. .- . .< m --

TABLE OF CONTENTS Section Papge 4.4.2 STRUCTURAL INTEGRITY 4-35 4.4.3 HYDROGEN PURGE SYSTEM 4-37 4.5 EMERGENCY LOADING SEOUENCE AND POWER TRANSFER, EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4-39 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 4.5.2 EMERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEAT REMOVAL SYSTEM LEAKAGE 4-45 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL R0D DRIVE SYSTEM FUNCTIONAL TESTS 4-48 4.7.2 CONTROL ROD PROGRAM VERIFICATION 4-50 4.8 MAIN STEAM ISOLATION VALVES 4-51 4.9 EMERGENCY FEEDWATER PUMPS PERIODIC TESTING 4-52 4.9.1 TEST 4-52 4.9.2 ACCEPTANCE CRITERIA 4-52 4.10 REACTIVITY ANOMALIES 4-53 4.11 INTENTIONALLY BLANK 4-54 l 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM 4-55 4.12.2 REACTOR BUILDING PURGE AIR TREAIMENT SYSTEM 4-55b 4.12.3 AUXILIARY AND FUEL HANDLING EXHAUST AIR TREATMENT SYSTEM 4-55d 4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 REACTOR BUILDING PURGE EXHAUST SYSTEM 4-57 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS (SNUBBERS)_ 4-60 4.18 FIRE PROTECTION SYSTEMS 4-72 4.18.1 FIRE PROTECTION INSTRUMENTS 4-72 4.18.2 FIRE SUPPRESSION WATER SYSTEM 4-73 4.18.3 JELUGE/ SPRINKLER SYSTEM 4-74 4.18.4 CO2 SYSTEM 4-74 4.18.5 HALON SYSTEMS 4-75

~

4.18.6 HOSE STATIONS 4-76 4.19 OTSG TUBE INSERVICE INSPECTION 4-77 4.19.1 STEAM GENERATOR SAMPLE SELECTION AND INSPECTION METHODS 4-77 4.19.2 STEAM GENERATOR TUBE SAMPLE SELECTION AND INSPECIION 4-77 4.19.3 INSPECTION FREQUENCIES 4-79 4.19.4 ACCEPTANCE CRITERIA 4-80 4.19.5 REPORTS 4-81 4.20 REAcr0R BUILDING AIR UtEATMENT 4-86 5 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATICN SYSTEM 5-3 5.3 REACTOR 5 '4 5.3.1 'LF. ACTOR CORE 5-4 5.3.2 ftEACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 NEW FUEL STOPAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 111

TABLE 4.1-3 MINIMUM SAMPLING FREQUENCY Item Check Frequency

1. Reactor Coolant a. Radio-Chemical Analysis (1) Monthly
b. E determination (2) Semi-annually
c. 15 Min. Gross Degassed 5 times / week when Tavg Beta-Camma Activity (1) is greater than 2000F
d. Tritium Radioactivity Monthly
e. Chemistry (C1, F and 0 2) 5 times / week when Tavg is greater than 2000F
f. Boron Concentration 2 times / week
2. Borated Water Storage Baron Concentration Weekly and after each Tank Water Sample makeup when reactor coolant system pressure is greater than 300 psig or Tavg is greater than 200 F
3. Core Flooding Tank Boron Concentration Monthly and after each Water Sample makeup when RCS pressure is greater than 700 psig
4. Spent Fuel Pool Boron Concentration Monthly and after each Water Sample makeup
5. Secondary Coolant _a. 15 Min. Gross Degassed Weekly when reactor Beta-Gamma Activity coolant system pressure is greater than 300 psig or Tavg is greater than 2000F
b. Iodine Analysis (3)
6. Boric Acid Mix Tank Boron Concentration Twice weekly (4) l or Reclaimed Boric l

Acid Tank l

10. Sodium Hydroxide Concentration Quarterly and after Tank each makeup j 11. Sodium Thiosulphate Concentration Quarterly and after j Tank each makeup
12. Condenser Partition 1131 Partition Factor Once if primary / secondary
Factor leakage developes, i.e.

( Gross Beta-Gamma on secondary side of OTSG

' 1s greater than 2 x 10-8 microcuries per cc and evidence of fission products is present 4-9 l

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(1) When radioactivity level is greater than 10 percent of the limits of Specification 3.1.4, the sampling frequency shall be increased to a minimum of 5 times per week.

(2) 5 determination will be started when the 15 minute gross degassed beta-gamma activity analysis indicates greater than 10 pCi/ml and will be redetermined each 10 uCi/ml increase in the 15 minute gross ' degassed beta-gamma activity analysis. A radio chemical analysis for this purpose shall consist of a quantitative measurement of 95 percent of radionuclides in reactor coolant with half lives of 2 30 minutes.

(3) When the 15 minute gross degassed activity increases by a factor of two above background, an iodine analysis will be made and performed thereafter when the 15 minute gross degassed beta-gamma activity increases by 10 percent.

(4) The surveillance of either the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank is not necessary when that respective tank is empty.

l 4-10

4.4.2 Structural' Integrity Specification 4.4.2.1 Inservice Tendon Surveillance Requirements The surveillance program _for structural integrity and corrosion pro-tection_ conforms to the recommendations of the U. S. Atomic Energy Commission Regulatory Guide l.35, proposed Revision 3, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment

Structures." The detailed surveillance program for~the prestressing system tendons shall be based on periodic inspection and mechanical

. tests to be performed on selected tendons, as specified hereafter.

4.4.2.1.1 Containment Tendons Tendon surveillance was completed for one, three and five years following initial structural: integrity using a Tech. Spec. based on Regulatory Guide 1.35 Rev. 1.1 The containment tendon structural integrity shall be demonstrated at five year intervals thereafter by:

a. Determining that for a representative sample
  • of at least 23 tendons (6 dome, 7 vertical, and 10 hoop)-each tendon has a lift off force equalling, or exceeding, its lower limit predicted for the time of the test as defined in NRC Regulatory Guide 1.35,

" Inservice Inspection for Ungrouted Tendons in Prestressed Concrete Containments", Proposed Revision 3, April, 1979.

If the lift off force of a selected tendon in a group lies between the prescribed lower limit and 90% of that limit, one tendon on each side of this tendon shall be checked for their lift off forces.

If the lif t off forces of the adjacent tendons are equal to, or -

greater than, their prescribed lower limits at the time of the test, the single deficiency shall be considered unique ~and acceptable. If the lif t off force of either adjacent tendons lies below the prescribed lower limit for that. tendon, the condition is report-able per T.S. 6.9.2.A3.

If the lift off force of any one tendon lies below 90% of its pre-scribed lower limit, the tendon shall be considered :a defective tendon. It shall be completely detensioned and a determination made as to the cause of the occurrence. The condition is reportable per T.S. 6.9.2.A3.

If the inspections performed at one, three, and five years indicate no abnormal degradation of'the pqst-censioning system, the number of tendons checked for lift off force during subsequent tests may be reduced to a representative sample of at least 11 tendons (3 dome, 3 vertical, and 5 hoop).

  • For each inspection, the tendons shall be selected on a random but representative basis so that the sample group will change somewhat for each inspection; however, to develop a history of tendon performance and to correlate the observed data one tendon from each group (dome, vertical, and hoop) may be kept unchanged af ter the initial selection.

4-35

4.8 MAIN STEAM ISOLATION VALVES Applicability Applies to the periodic testing of the main steam isolation valves.

Objective To specify the minimum frequency and type of tests to be applied to the main steam isolation valves.

Specification 4.8.1 A check of valve stem movement, up to 10 percent, shall be performed on a monthly basis when the unit is operational and under normal flow and load conditions.

4.8.2 The main steam isolation valves shall be tested at latervals not to exceed the normal refueling outage. Closure time of

< 120 seconds shall be verified. This test will be performed (

under no flow and no load conditions.

Bases Since a portion of the main steam lines and the steam lines to the main feed pump turbines are located in the turbine hall which is not protected against hypothetical tornado, missile, or aircraf t incident; main steam isolation stop check valves are provided and located in the hardened portion of the intermediate building. These stop . check valves are remotelyclosedbytheoperatorfroyghecontrolroom,closeinlessthan two minutes, and are tight closing for long term containment isolation.

Their ability to close upon signal should be verified at intervals not to exceed each scheduled refueling shutdown, and valve stem freedom should be checked on a monthly basis.

References (1) FSAR, Section 10.2.1.3 I

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