ML17195A001
| ML17195A001 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/03/1983 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17195A002 | List: |
| References | |
| NUDOCS 8311070290 | |
| Download: ML17195A001 (11) | |
Text
e UNITED STATES e
NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 COMMONWEALTH EDISON COMPANY DRESDEN NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO. 50-237 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 77 License No. DPR-19
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated May 24, 1978, as supplemented by letters dated July 15, 1981 and May 3, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application*,
the provisions of the Act arid 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 health and safety of the public, and (ii) that such activities will be conducted in comp 1 iance with the Cammi SS ion Is regulations; D.
The issuance of this amendment will not be inimical 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 Commission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Provisional Operating License No. DPR-19 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 77, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance.
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 3, 1983 FOR THE NUCLEAR REGULATORY COMMISSION L~~~:d Dennis M:'cru~~~!
Operating Reactors Branch #5 Division of Licensing
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 COMMONWEALTH EDISON COMPANY DRESDEN NUCLEAR POWER STATION, UNIT* NO. 3 DOCKET NO. 50-249 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 68 License No. DPR-25 L
The Nuclear Regulatory Commission (the Commission) has found that:.
A.
The application for amendment by Commonwealth Edison Company.
(the licensee) dated May 24, 1978, as supplemented by letters dated July 15, 1981 and May 3, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions 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 health and safety of the public, and (ii) that such activities will :be conducted in compliance with the Commission's regulations; *
' l.
D.
The issuance of this amendment will not be inimical to the common*
defense and security or to, the hea 1th and safety of :the public; and
~-
E *. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have
- been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.8 of Facility Operating License No. DPR-25 is* hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 68, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance.
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 3, 1983 FOR THE NUCLEAR REGULATORY COMMISSION L~~~~
Dennis M. Crutchfield Chief Operating Reactors Branch #5 Division of Licensing
ATTACHMENT TO LICENSE AMENDMENT NO. 77*
PROVISIONAL OPERATING LICENSE NO. DPR-19 AND AMENDMENT NO. 68* TO FACILITY OPERATING LICENSE NO. DPR-25 DOCKET NOS. 50-237/249 Replace page 56 of the Appendix A Technical Specifications with the enclosed page 56.
This revised page contains the captioned amendment number and a vertical line indicating the change.
- During the issuance of License Amendment No. 76 to DPR-19 and License Amendment No. 67 to DPR-25 a date and the amendment nos. were inadvertently omitted from pages 9lb and 99b. Therefore, corrected pages are attached hereto.
3.3 LIMITING CONDITION FOR OP~RA'flON B.
Control Rods I *
- 2.
All control rods shall be coupled to their drive mechanisms when the mode switch is in "Startup" or* "Run".
With a control rod not coupled to its associated drive mechanism, operation may continue provided:
- a.
Below 20% power, the ~od shall be declured inoperable, full inserted, and the directional control valves electri-cally disarmed until recoupling con be attempted at all-rods-in or at power levels above 20 percent power.
- b.
Above 20 perc.ent power, recoup ling is being attempted -in accordance with an established procedure or the rod shall be declared inoperable, fully inserted and the directional control val~es elec-trically disarmed.
The control rod drive housing-support system shall be in place during reactor power operation ahd when the reactor coolant system is pressurized above atmos-pheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification J.J.A.I is met.
Unit 2 Amendment No.
Unit 3 Amendment No.fl,
/ '
- 4. 3 SURVEILLANCE REQUllU-:MENT B.
- Coupling Integrity
- a. The coupling integrity of each control rod shall be demonstrated by withdrawing each control rod to the fully withdrawn position and verifying that the rod does not go to the overtravel position; (I) Prior to reactor criticality after completing alteration of the renctor
- core, (2)
Anytime the control rod is withdrawn to the "Full out" position in subse-quent operation, and I
(3)
For specifically affected individual control rode *following maintenance on or modification to the control rod or rod drive system which could affect the rod drive coupling integrity.
b, Normal operating ~roctice is to observe the expected response of ~he nuclear instrumentation to verify that the con-trol rod is following its drive each time that control rod is withdrawn.
For control rod drives that have experienced uncoupling and no response is discernable on the nuc-lear instrumentation, the response should be verified ~hen the reactor is operating at power levels above.20 percent.
- 2.
The control rod drivehouaing nupport system shall be inspected after reassembly and the results of the inspection recorded.
56
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T.-6 LIMITING CONOITION FOifOPERATION
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I. Snubbers (Shock Suppressors)
- 1. During all modes of operation except cold shut-down and refuel, all safety related snubbers limited in Table 3.6.l~ and 3.6.lb shall be aper~
able except as noted in Specification 3.6.1.2
,through J.G.1.* 4.
- 2.
Fr0m and 'after the time a snubber is determined to be inoperable, continued reactctr operation is pennis-sib le only during the succeedlng 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless the snubber is sooner made operable or replaced. Torus Ring Header snubbers may be inoperable in either of the following configurations until January 19, 1984,.
to facilitate the installation of the Mark I torus attached piping 111.?difications! ____, _________.,...... ___....... _
Confiauratlon Aa l**Sl' other **l*tloa'anubb*r palr (up to 3 palra) on tb.* ECCi budn, or Cooflaur*tloo 11 On* **l1tln1 *nubber frOlll **ch of the 6 exiatioa anubber palra on the ICCI he*Jer.
- 3. If the requlremeoC. of 3.6.1.1.and 3.6.1.2 cannot be IDOt, *o orderly ehutdovn ehall be lPlthted and tba reactor *hall be in cold *hutdOWD or refuel condldoa vltbin 36 houra.
- 4. If a 1oubber la deteTialoed to ba looporable vhlle th*
reactor l* lo th* cold 1butdowo or refuel 1110de, th*
anubber *hall ha aade op*rabla or replaced prior to reactor atartup. Thia requlre1110ot doee not apply to toru1 Ring lluJar anubbera for tho~ period ldentUlacl ln peragraph 3.6.I.2 above *.
- l. Snubber* ma7 ba added to aafet7 related ayato~ vltb-out prior llceoaa amendment to Tablea 3.6.la..ud/or 3.6.lb provided tb*t a re*laloo to Tablaa 3.6.la and/or 3.6*1b la locludacl vltb th* next llcen** amand-
-nt r*llue*t.
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iJnuhhoru (Uhuclt Dupprooooro)
'l'ho fo llnwlnl~ oaarvo i llnncc roqu ! rc1111:nto O(**p ly to Lt unfcty rolut:u*I onulibora l!uloJ ln 1'uhlco l *. 6.ln nr l.6.lh.
- 1. Vlouol Jnovcc~lon.
An l1Hlcpon1lont vlsunl lnopcc;~lon ohol l ho pcr-c fonncd un Uao aqfcty rclolecl hy1h-oulic on,l 1noch11nlcnl onubl>cru coolninotl in 1'11hloo J.6 *. **
nt\\ll J.6.11, ln accordunco with ll11!. hulou "ct1ul ft
- A 11 h y cl r nu li c on uh b t! u
\\./hon c!
111? 11 I 111111 u d n l hno Iman domonutr~al:',HI by opcrnl int; xpcr-iencu, lnlJ teat lnc 01* una I y11 iu tu lie com-pnt ih lo with the upcrutins c11vin*11111c11l.
1hnll \\Jo v1ouolly lnope!cl:ccl.
'fhl.o l11npcc-t lon ohnll 1ncludu 1 liul nol nccc!1111n1*l ly ho limilccl to, lnupcolion o( lhc hych*oullr.
flulcl rcnnrvoir, fluid co1111cctio11a, tlncl llnkllno connection to Lhn pipi11r. 111111 n11cho*
r.o vodfy onulJhor opornhilily.
b, All 111cchanlcnl snuhltcronhnll lie vi111111lly i1101>octtid.
Thie inopcclion vlrnll cnn11lo1~
of, but not ncccoondly l*c I i111il,!1I to.mil**
aper.lion of. tho u11uhlrn1* 111111 11ll.11cl1111r.111'i'Lo lht! l'fpinn and onchor for JnilJc:nl iouu o{
dL1mnuo ol" laipoh-<!d opornblllt-y.'
llo, of Gnuhhors found lnnpernhlu J>urlnc I nu pee t inaLJ.n.l crvn 1
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2 JI It
.5, 6, 1
~u I
tluxl llc!*111l1*1Hl
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~.!'.Y !!!t lU 111011l.h0., 2'.il 11 lllllll l h II
- 7 J l; 6 lllflUlho I 1'.il 1211 1l11y11 1 ~* :t
- ()l cJ II y 11 f 1Yl JI cl 11 y u
- nl: ~
The inspecti\\'.lrl frequency is bqQe results require a shorter inspe~C:ion interval will over~ide the previous schedule.
To further increase the assurance of snubber re-liability, functional tests wi11 be p~rformed once each refueling cycle.
A represent4C:ivc sample of 10% of the safety-related snubb~rs will be func-tionally tested.
Observed failures on these samples*
will require testing of additional uni~s.
Hydraulic snubbers and mechanical anubbers may each be treated es different entities for l:he above surveillance programs.
Hydraulic snubber testing will include stroking of the snubbers to verify piston movement, lock-up, and bleed.
Functional testing of the mechanical snubbcrs will consist of verifico~iun that the force that initiates free movement of the snubber in either tension or ~ompression is icss than the maximum breakaway friction force, The remaining portion of the functional test cons~sting of veri-fication that the activation {restraining action) is achieved within the specified r~nge of accelera-tion in both tension and compressioq will not be done.
This is due to the lack of competitive marketable test fixtures available for station use.
Therefore, until such time as test fixtures become available, only port (i) of the 'test will be per-formed; port (ii) will not be done.
Unit 2-Amendment N~.;ur-: 76 Unit 3 -Amendment Nos.ft~, 67.
DI? it*.1 ~,.*
and DPR-25
~1en the cause of rejection of the snubber is clearly established and remedied for that onubb'er and for ony other snubbers that may be genericol~y
- susceptible, and* verified l>y inscrvice f unc t ioru~ l testing, that enubl>er may be exempted from l>eing counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have ~he same design features directly related to rejection of the snubber by visual in-spection or are similarly located or exposed to the same environmental conditions such as tempera-ture, radiation, and vibration.
Monitoring of snubber service life shall ~onsist of the existing station record systems, including e the central filing sya tem, maintenance files, sqfety-related work packages, and snubber inspection records.
The record retention programs employed at the station shall allow etation personnel to main-tain snubber integrity.
The service life for hydraulic snubbers is 10 years.
The hydraulic snubbere existing locations do not impose un~ue safety implications on the piping and compone11ts be-cause they ore not exposed to excesses *in environ-mental conditions.
The service life for mechanical snubbers ie 40 years, lifetime of the plant.
The mechanical snubbers are installed in areas of harsh environmental conditions because of their dependa-bility over hy<Jraulic s.nubbera in these areas. All snubber inetallat ions have been tt,oroughly engineer..
providing the necessary.~afety requirements *. Evolu
- ~
tions of oll snubber locations and environmental
"-._./
conditions justify the above conservati.vc enullber servlc,e 1 i ves.
- A re-analysis of the ring header design based upon acceleration
. response spectra derived from the original suction header analysis
. :*report demonstrates that for nonnal operation plug seismic, neither the header nor the torus penetration are over-stressed with all
The limitation of a'maximum of 3 pairs or 1
. snubber from each pair inoperable out of 6 pairs is considered
'conservative. Since the analysis shows that the plant can operate
.
- tion on operation and startup with inoperable snubbers until
- January 19, 1984 is justified. This time frame is adequate to allow completion of the Mark I torus attached piping modification.
99b
e UNITED STATES e
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
-SUPPORTING AMENDMENT NO. 77 TO PROVISIONAL OPERATING LICENSE NO. DPR-19
~ ' {: **~*
AND AMENDMENT NO. 68 TO FACILITY OPERATING LICENSE NO. DPR-25 COMMONWEALTH EDISON COMPANY
~ -_.
1.0 INTRODUCTION
DRESDEN NUCLEAR POWER STATION, UNIT NOS. 2 AND 3 DOCKET NOS. 50-237 AND 50-249 By letter dated May 24, 1978, as supplemented July 15, 1981, and May 2, 1983, Commonwealth Edison Company (CECo) (the licensee) proposed amendments to Appendix A of Operating License Nos. DPR-19 and DPR-25.
The subject change involves Section 4.3.B.l.b of the Technical Specifications for Dresden Unit
-Nos. 2 and 3. The licensee has proposed to amend Section 4.3.B.l.b, -
Surveillance Requirements for Control Rod Coupl.ing Integrity, to modify the wording for better understanding.
A Notice of Consideration of Issuance of Aniendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing. related to the requested action was published in the Federal Register on September 21, 1983 (48 FR 43131).
No request for hearing was received and*
no comments were received.
_ 2.O EVALUATION Because of an early history of occasional control rod uncoupling in the Dresden reactors, Technical Specification surveillance for uncoupling verifications in addition to those now required in the Standard Technical Specifications for General Electric Boiling Water Reactors, NUREG~0123, -
were included as specification 4.3.B.l.b. This required confirmation of coupling using nuclear instrument response during a rod notch withdrawal *. - *
- When no instrument re~ponse was discernable at lower power, the response should be verified when the reactor is operating at power levels about 20%.
The intent of the specification was to provide a general check for all -
control rods but specifically for those rods with an uncoupling history.
The existing Technical Specifications 4.3.B.1.b wording implies that all rods be reverified at power levels above 20%.
To clarify the intent of surveillance requirement 4.3.B.1.b, CECo proposed a wording change in a letter dated May 24, 1978.
In a letter dated May 2, 1983, the wording was modified to further clarify the intent and consistency.
with the description and analysis in the Final Safety Analysis Report (FSAR).
The Final Safety Analysis Report {FSAR) Section 3.5.4, Surveillance and Testing for Control Rods, states in part " *** During reactor operation individual control rod drive mechanisms can be actuated to demonstrate functional performance.
Each time a control rod is withdrawn a notch, the operator will observe the in-core monitors' indications to verify that the control rod is following the drive mechanism.
When *the operator withdraws a control rod full out of the core, he tests the coupling integrity by trying to withdraw the rod drive mechanism to the overtravel position. Failure of the drive to overtravel demonstrates rod to drive coupling integrity."
FSAR, Section 6.5.1, Design Basis for Control Rod Velocity Limiters, states in part " *** The purpose of the control rod velocity limiter is to reduce the consequences in the event a high-worth control rod became detached from its rod drive and dropped out of the reactor core.
11 FSAR Section 14.2.1, Control Rod Drop Accident, shows that the analysis is based upon a fully inserted control rod assumed to fall out of the core after becoming disconnected from its drive and after the drive has been removed to the fully withdrawn position.
In order to assure that the control rod remains connected to its drive, and in the *interest of good operating practices, the licensee's proposed change to Technical Specification 4.3.B.1.b,
. reaffirms the FSAR.
Further, when the nuclear instrumentation does not pro-vide evidence of the control rod movement, e.g., during a startup, the proof of coupling integrity for rods with uncoupling history will be conducted at a power level in excess of 20% where local power range monitors will give the necessary indication.
In addition, the licensee submittal of July 15, 1981 references General Electric Service Information Letter, SIL #52, Supplement 2, July 31, 1974, which shows that improper installation of the control rod drive inner filter had resulted in causing control rods to become uncoupled when *they reach position 48 (fully withdrawn).
The licensee has implemented the improved GE overhaul procedure and test to assure proper installation of the inner filter. This has resulted in a significant reduction in events of uncoupled control rods.
Based on the foregoing, the staff finds the licensee's proposal to improve the wording of Technical Specification 4.3.B.1.b to be acceptable.
3.0 ENVIRONMENTAL QUALIFICATION The staff has determined that the amendments do not ~uthorize a change in effluent types or to ta 1 amounts nor an increase in power 1eve1 and wi 11 not result in any significant environmental impact.
Having made this deter-mination, the staff further concludes that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 51.5.(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal, need not be prepared in connection with the issuance of these amendments.
.*.-"*~*.. -.. -
4.0 CONCLUSION The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ACKNOWLEDGEMENT The following staff members contributed to this evaluation:
T. M. Tongue K. R. Ridgway T. N. rambling Dated:
November 3, 1983
_. **:.