ML19343C604
| ML19343C604 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 02/26/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19343C599 | List: |
| References | |
| NUDOCS 8103240582 | |
| Download: ML19343C604 (13) | |
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UNITED STATES i
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NUCLEAR REGULATORY COMMISSION 3 i
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CONNECTICUT YANKEE ATCMIC PCWER COMPANY DOCKET MO. 50-213 RADCAM NECK PLANT AMENDMENT TO FACILITY OPERATINS LICENSE Amendment No. 37 License No. DPR-61 1.
The Nuclear Regulatory Concission (tne Concission) has found that:
A.
The application for amendment by Connecticu Yankee
-Atomic Power Cc any (the licensee) dated June 29, 1977, as revised by letters dated May 26,1975, May 29,1978, July la,1975, April 27,1979, June 29,1979, March 25, 1930, and November a,1950, and the a::lica:icn for amendment dated.Nevemoer a,1980, ccmolies with the standards and recuirements of :ne Atemic Energy Act of 195*, as amended (the Act), and the Commission's rules and regula: ice.3 set for:n in 10 CFR Cha::er I; E.
The facility will crerate in cenformity witn the application, the Orovisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable: assurance'(i) that the activities authorized by this amendment can be conducted without endangering the healtn and safety of the cu:lic, and (ii) that.such activities will be conductec in compli-ance with the Coemission's regulations; D.
The issuance of this amendment will not be inicical.to the~ common defense and security or to the health and safety of the public~; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Concission's regulations and all applicable recuirements have been satisfied.
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2-2.
Accordingly, the license is amanded by changes to the Tecnnical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C(2) of the Facility Operating License No. DPR-61 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specificaticns contained in Apsendices A and B, as revised through Amendment Nc. 37, are hereby incorporated in the license. Tne licensee shall operate the facility in accordance witn the Technical Specifications.
3.
This license amendment is effective as of :ne cate of its issuance.
FOR TFE NUCLEAR REGULATORY COMMISSION Qf Dennis M. Crutchfield, C...ef Operating Reactors Branen #5 Division of Li:ensing Attacnment:
Chances to tne.Tecnnical Specifications Date of Issuance:
February 26, 1981 9
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r ATTACHMENT TO LICENSE AMENDMENT fi]. 37 FACILITY OPERATING LICENSE NO. DPR-61 DOCKET NO. 50-213 Replace the following pages of Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES III III V
4-19 4-19 4-20 4-20 4-21 4-21 Table 4.10.1 4-22 c 22 3 20 3 25 3-26
-3 26 4-4 4.c a-La L.:a
- This page is Lincluded forLthe purpose of reflecting the Environmental.
-Qualification provisions (Section 6.14) issued by NRC Order detec October 24, 1980.
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SECTION 4.0 SURVEILLANCE REQt.IREMENTS PAGE 4.1 Introduction 4-1 4.2 Operational Safety Items 4-2 4.3 Core Cooling Systems - P,eriodic Testing 4-3 4.4 Containment Testing 4-5 4.5 Emergency Power System Periodic Testing 4-11 4.6 Radioactive Liquid Waste Release 4-13 4.7 Radioactive Gaseous Waste Release 4-15 4.8 Auxiliary Steam Generator Feed Pump 4-17 4.9 Main Steam Isolation Valves 4-18 4.10 Inservice Inspection and Reactor Vessel Surveillance 4-19 4.11 Deleted 4.12- 'High Energy Piping System Tests 4-24 4.13 Hydraulic Snubbers 4-25 4.14 Flood P otection ' Annunciation 4-29a 4.15 Fire Protection S'ystems 4-30 SECTION 5.0 DESIGN FEATURES 5.1 Introduction 5-1 5.2 Site Description 5-1 5.3 Reactor 5-2 5.4 Containment 5-4 i
III Amendment No. 24, 27, 5
INDEX Ar*.INISTRATIVE CG*iTROLS FAGE 6.0 ADMINISTRATIVE CONTROLS SeCTIOn Er9 Co n s ul ta n ts..........................................
6.5.2.4 E"9 6.5.2.5 Yeeting ' Frequency............
E-9 6.5.2.6 Quorum,,...........................................
5-10 6.5.2.7 Review...............................................
5411 Audits...,...........................................
6.5.2.8
'E-11 Authority............................................
5-12 6.5.2.9 6.5.2.10 Records..............................................
f-12 6.5.2.11 Qu al i f i c a t i o n s........................................
6 12 6.6 REPORTABLE OCCURRENCE ACTION..:.............................
6-13 6.7 S AFETY LIMIT VIOLATION......................................
$_g3 6.8 PROCEDURES......................................
6.9 REPORTING REDOIRE.v5. hts
$_15 6.9.1 R;3 TINE REPORTS...............................,,,,,,,,,,,,
6-17
- '** M LE *~*IUINCES *
- 6-22 6.9.3 S?hCIAL REP 0KTS...........................................
6-22 RECORD RETENTION..........................................
6.10 6-2.,.
sin. ION. P.0.-...... P,s0.
.v...............................
s 3 :v. 2 v..
ess 6.11 b.%
6.12 RESPIRATORY PROTECTION PROGRAM _
6.12.1 ALLOWANCE........J..........................'.....******
6-24 6-24 6.12.2 P ROT E CT IO N P ROGRAM....................................
6-26 6.12.3 REVOCATION..............................................
HIGH RADIATION AREA................,..................... '6-26 6.13 ENVIRONMENTAL C'JAL IF IC ATION........
6.14 1
V Amendment No. JJ, 27
3.14 ?RIMARY SYSTEM LEAL *JGE_
AF?LICASILITY:
Applies to limiting operation of tha plant under varying ratec and cend'.tions of pri=ary plan: leakage.
03.T TVE:
To specify pri=ary plant operability based upon pri=ary plant leaka58-SPECIy1CA* IONS:
A.
Operation of the reactor coolant syste= shall be per=1tted ade the followint leakage criteria.
1.
One GPM unidentified and uncontained leakage in the reazar coolan: syste=.
2.
Ten G?M in the reae:c: coolant system 3.
Six liter? /hr. from residual heat re:noval syst.c=
seals, flanges, and valves.
4.
No pressure bounda y leakage allowed in the reactor co: Tac system.
5.
?:i=ary - :o - Secondary leakage thrcugh the stea=
generato: :ubes shall be li=1:ed to 0.4 G?M :otal for all s:ca= genera: ors =c isolated from the reactor coolan: sys:e= and 150 gallons per day tL ough any one staan generato ne: iselated from the reactor coolant syste=.
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6.
Leakage throuch each of the followi' ; ECCS valves V
shall not excee'd one GOM:
SI-CV-Ef 2.'. - SI-CV-8625 -
SI-CV-862C - SI-CV-S52D - CD-CV-87"A - CD-CV-8725 V
3.
ACTICS ;5 72 QUIRED USOER ~?.E FOLLOWING
- NO!!! CSS:
1.
Ki:h any ??2SSURE 300'* ARY LEAKAOI, be in COLD SE'.'. DOWN vithin :he f:llowing 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
. With any Esac:or Cocian: Syste:. leak. age greate:
-br.. any one of :he above lici:s, excluding PRESSURI EOUNDARY LEAEj.GI, redu:e the leakage ra:e within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> c he in COLD SE7TD025 within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
3.
If primary o secondary tube leakage (not including leaks originating fro = tube to tube sheet welds) in excess,of 'specificarica 3.14.A.5 above', results in a
. cold shu:down, inservice inspections shall be performed in acec: dance with the first sa=ple inspection in Table L.10.1-2 during that cold shutdown.
Unc:ntained leakage 'is flov := any open system. This includes I A.5:S :
leakage to the contain=en: and pri=ary auxiliary building su=ps.
Leakage which is attributed :o a specific component is considaad to be Lian:ified. Unidentified leakage would be that flow whout
. path is no: known.
Leakage f::= the pri=ary plan: can be detected in a nu=her of ways including centainnen: sc=p level, contain=en: hu=idity a=E a ; particula:e =easurece.nts, =ain:enance of water volume inventories and routine surveillance of charging header flow.
- Leakage that is both uncontained and - unidentified is 'undesirallie
. Amendment No. 37 f c= the point cf safety. A leak rate of one gpe can be dent vithin a nu=her of hours and without cause or definition the 4
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plan: thould be shutdown.
Uncontained but identified leakage does not constitir.e a safety hazard if it can be determined tha: operations can safely continue. Kneeledge of the source and path of the Jaak pereits a sound judge =ent to be made regarding continued plaze operation.
Ten gp= is well within the =1:1=u= =ake-up capabd i'-**a but 1: is desirable to initiate plant shutdcen.
Leakage that is both contair.ed and identified does at constitute a safety hanard if it can be dete =ined that plant speratio=s can be safely continued. 10G'?Mleakage is well with*a the capacity of one charging pu=p (360 sp=) and makeup immuld be a.ailable even under the coinciden: less of offsite power condi: ion.
Centaining the leak within other auxiliary syste=s pe:=its control ever dispesitien of the va:ar volu=e and at:ivity e=anating frc= the.pri=ary system.
The plan: is expected te be opera ed in a =anner such that the secondary cociant vill be =aintained vithin those chemistry li=its found to resui in negligible cc :csien of the stem generator tubes.
If the secondary coolan: che=1stry is no =aintained within these li=1:s, 1ccali:ed corresien =ay likely result in stress corrosion
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cracking. Tne exten cf cracking during plan: ope =stion vould be li=i:ed by the li=1:a: ion of stes= genera:er :ube leakage be:veen
- he pri=a-v coelant sys:e= and the secondary coolazz syste=
(primary-:c-secondary leakage = 150 gallens per day per stea=
genera:o:).
Cracks having a pri=ary-:c-secendary 2eakage less than this li=1: during eperatien vill have an adequa:e =argin ef
, saf ety :c vi:hstand :he leads i=p: sed during-ner al operation and by pcstulated accidents. Operating plan-s have fe=cestrated that pri=ary-:e-se:endary leakage of 15C gallens per day per stea: genera::: can readily be de:e::ed by radiatine ment:oring cf stea= genera:c: blevdern. i.eakage in excess of this lini:
vill require plan: shutdei.-. during v'..ich the leaking tubes vill be located and plugged.
Leakage from pump seals in the residual heat removal pumps of eight liters / hour is considered as indication that the seal has failed. Six liters / hour is considered as an indication of im-pending failure. The radiological hazard associated with six liters / hour leakage from the residual heat removal system is acceptable. The two-hour dose to the thyroid at the site bounury, folicwing the maximum hypothetical accident, with maximum cllowable containment leak rate, would be increased from 60 to 90 rem.
Excessive leakage through certain ECCS check valves could indicate that the valves are not performing their function of preventing reverse flow. The. configuration of these valves is such that their failure to ' function could result in an
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intersystem loss-of-coolant accident.
Reference:
FDS A; Section 5.2.1 FDSA; Section 10.4.4 AmendmentNo.k.37 3-25
B)
During nor=al operating psriods, c =anuni test of all actuated co=ponents shall b2 conducted to de=on-strate operability. The tent shall be perfor=ed in accordance with written procedure as r--->:ized below:
1)
Monthly, each of the high pressure saf ety inj ection pu=ps, each of the low pressure saf ety inj ection (core deluge)'pu=ps, and each of the residual heat re= oval (RER) pu=ps shall be individually test run on recirculation.
2)
Monthly, the charging pu=ps and =etering pu=p shall be tidividually test run.
3)
A: cold shutdo n.
, all safety inj ectico and core deluge valves vill be cycled under "no-flow" conditions.
C)
If one of the high pressure saf e:y inj ection pu=ps or one of the low p'ressure safe:y injee:ien (Core deluge) or one of the residual heat re= oval (EE?.) pu=ps is cut of service, the re=aining pu=p shall be tested within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at subsequent intervals cf no: greater :han 7 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> s.
D)
During each refueling shutdown, the re=otely controlled,
=otor-operated contain=en: spray wate: valve shall be operat ed under a "no-flow" condition. The test vill be considered satisfactory if visual observation shows that the valve has operated satisf ae:::ily.
I)
One ce trifugal charging pe=p and the E?S1 pu=ps shall be de= ens::a:ed inoperable at least once per 31 days when-ever the te=perature of one or ::re cf the non-isolated RCS cold legs is less than c: equal :: 340 F and the RCS is not ven:ed by a =ini=u= epening of 3 inches, by verify-ing tha: the charging pu=p cont :1 switch is in the trip pullout position and red tagged and that the E?S1 breaker cabine:s are locked and tagged cut and the E?SI pe=p discharge valves are locked closed.
F)
Periodic leakage testing of each ECCS ct ck valve listed in Table 4.3.1 shall be accomplished prior to entering operational mode 1:
- 1) After every time the plant is placed in the cold shutdown condition for refueling.
- 2) After every time the plant has been placed in the cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months.
- 3) Prior to returning the valve to service after main-tenance, repair, or replacement work is performed.
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4.a Amendment No. }f, 37
Leakage may be measured indirectly using pressure indicators, if accomplished in.accordance with approved procedures and supported by computations showing that the method is capable of demonstrating compliance with the leakage criteria of section 3.14.A.6.
The minimum differential pressure across these check valves during these leakage tests shall not be less than 150 psid.
e Table 4.3.1 - ECCS Check Valve l
SI-CV-862A SI-CV-862B SI-CV-862 C M hl h n Ohk gKb SI-CV-862 D g Ul\\n CD-CV-872A CD-CV-872B U
Easis:
The core cooling syste=s are the prised 7al plant saf eguard.
They provide the = cans to insert negate e reactivity and limit core da= age. in the event of a loss-of colar" incid ent.
Pre-operational perfor=ance tests of the ce=ponents are perfor=ed in the =anuf acturer's shop. As Laitial syste= flow test de=en-s :ates proper dyna =ic functioning of the systa=.
Thereaf t er, periedic tests de=enstrate tha: all ec=ponents are functioning properly.
In _ order :o assure that a pressure transient eccurring during the tes:ing cf the HPSI pu=ps vill not exceed the pressure and te=pe. -
ature li=its of specification 3.4, there =ust be appropriate-relief paths available; this is provided fer in specification 4.3.A.4 The separation of e=ergency power syste=s and associated core cooling equip =en into two independent groupings per=1ts co=plete fune:1on testi=g of the individual syste=s and equip =ent.
=::
The ect:ai:::en: spray va:e: is previded, if re:ciref, by the lov pressure safe:y injee:icn pe=ps, which are aise pa:: ef the :::e deluge syste=.
- is ne: desirable to :es: she valve a: =en:h y in:erra* s since 1:
recuires closure cf a =anual valve in :ba spray header.
This valve =ust be ' closed to preven: initiation of spray when the me:or-operated spray valve is open since the residual hea: re=cvai sys:c= will always be pr essurized. ! Closure of the =anual valve is act desirable at po,wer and, :herefore, dic:a:es that the notor-cperated spray talve be tested a: refueling intervals o=ly.
The surveillance in Spe 'fication I) assures tha: :he
'd d e g cenditions for operation required'for lov te:pera:ure overpressurization protectien have been =e:.
The surveillance in Specification F) tests the ocerabability of check valves which act a primary coolant system pressure isa.lation valves and thereby reduces the potential for an intersystem loss of c olant accident.
1eference:
(1)
FOSA - Section 5.2.7 (2)
FCSA - Section 9.5
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- Amendment Nc. }{, jf, 37 e
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4.10 INSERVICE INSPECTION AND REACTOR VESSEL SURVEILLANCE Apolicability: Applies to the periodic inservice inspection anc testing of ASME Boiler. and Pressure Vessel Code Class 1, 2, and 3 equivalent c =ponents.
Objective: - To verify the structural integrity of the applicable components defined above.
Scecification:
A.
Inservice Inspecticn of ASME Code Class 1, Class 2, and Class 3 equivalent components shall be perforrod in accordance with Section XI of the ASME Boiler and Pressure Vessel Ccde and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), with-the exemptions and alternate inspections that have been approved by the NRC pursuant to 10 CFR 50, Secticn
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10.55a(g) (5) (1). These-exemotions and alternate inspections are included in the Inservice Inspe::icn Program.
B.
Inservice testing of ASME Code Class 1, Class 2, and Class 3 equivalent pumps and valves shall be cerformed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code anc ap:licable Addenda as repuired by 10 CFR 50, Section 50.55a(g), with the exemptions.and alternate testing that have been a:oreved by tne NRC pursuant to 10 CFR 50, Section 50.55a(g).(5) (i). These.exemptiens anc alternate-test,ing are included in the Inservice Inspe::icn Program.
C.
One reacter coolant'pu=p flywheel shall be examined visually and 100% volumetrically every 'c:her refueling.
D.
The removal plan for the' remaining reactor vessel surveillance capsules is scheduled as fellows:
Capsule Capsule Removal Time Predicted Excosure (EulMET)R/CM2 Type Ident.
Yea rs*
Cacsule Reactor Venel ID I
K 10 1.3x1019 1.0x10II II~
D 15 2.0x1019 1.6x10TI II E
Cd 3.3x10l9 2.6x10lf I
B Standby I.
-C-Standby I
g Standby
- Rem: val' tire is.in approximate-calendar years frc= the date cf
. initial-criticality (July 24,.1957) and intended to coincide udth
.the nearest refueling outage to that tice.
4-19 Arenue.:. No. 5, 37
l Basis 10 CFR 50.55a(g) sets forth the inspection requirements for nuclear power plant components by referencing the ASME Boiler and Pressure Vessel Code, Section.XI, " Rules for Inservice Inspection of Nuclear Power Plant Components" which defines the requirements for inspection of ASME Code Class 1, 2, and 3 equivalent components.
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- In regards to the reactor surveillance capsule program there are'only two capsules (Type II) containing soecimens of weld metal.remainir.g. Therefore, the first of these will not i
be removed until 10 years of operation have elapsed and the second until 25 years have elapsed. This will' provide the greatest amount _of useful information about irradiation effects on weld metal over the lifetime of the plant.
In addition, removal of the Type I capsule at the 15 year point will confirm the noted trends and will leave three Type I capsules in standby for special purposes.
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~4-20 Amendment No. 5, 36,37 v
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4*II ERIMARY SYSTEM HYDRO TESTS I
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