ML17201M300
| ML17201M300 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/19/1983 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Delgeorge L COMMONWEALTH EDISON CO. |
| Shared Package | |
| ML17201M294 | List: |
| References | |
| LSO5-83-01-025, LSO5-83-1-25, NUDOCS 8901300184 | |
| Download: ML17201M300 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION Docket No. 50-237 LS05-83-0l-025 Mr. L. DelGeorge Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690
Dear Mr. DelGeorge:
WASHINGTON. D. C. 20555 January 19, 1983
SUBJECT:
SEP INTEGRATED ASSESSMENT STATUS FOR THE DRESDEN NUCLEAR POWER STATION, UNIT 2 By letter dated December 6, 1982, you provided your position on t~e 34 open topics summarized in Chapter 4 of the draft Integrated Plant Safety Assessment Report (IPSAR} for Dresden Unit 2.
In addition, the staff has received many other letters regarding specific topics over the past three months.
All of the infonnation you provided is being used to
- finalize the Dresden Unit 2 IPSAR.
- Publication of the* final report is scheduled for January 31, 1983.
The purpose of this Jetter is to provide you a status of all of the open items which will be identified in the final report *. contains a list of all the identified differences for which no further a~tion or backfit is required.
Included in this list are those topic~ where (1) the corrective action is complete, (2) the review is *tavered by ~nether NRC program, and (3) those items which were resolved in the draft IPSAR or as a result.of additional i nfonnati on
-received si;:ce -the draft IPSAR *
. Enclosure 2 lists those issues for which you have committed to implement
.hardware modifications to the facility.
For those items.identified in, the staff wi 11 re qui re a schedule. for:- canpl eti o.n of the modifications within 30 days of ~eceipt of this letter *.
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Mr. Those issues for which you have committed to.make procedural or technical specification (TS) changes are listed in Enclosure 3.
The procedural changes wi-ll be reviewed by the NRC Region III Office.
It is the staff's understanding ~Q.~,~.. ~.11.. of the procedural changes and the proposed TS changes will be ~6~pl~ted by the end of the current r~fueling outage.
If the actual schedule differs from our assumption, please infonn the staff within 30 days-of receipt of this letter. lists those *issues which require additional infonnation or analysis.
In some instances, you have provided further infonnation which is.und~rgoing staff review~.Those items are noted in the enclosure *.
However, the majority of the items are still outstanding.
In addition, you hav_e responded to a few i terns and the staff review found the infonna-ti on to be insufficient.
These items are described in Enclosure 5.
Of special note is that regarding SEP Topic *III-2, "Ventilation Stack."
As described in the enclosure, this issue is being reopened due to your eval-uation. being perfcinned using a m~thodology not accepted by the staff.
Please review the two enclosures and provide your schedules for completion of the required infonnation within 30 days of receipt of.this letter.
The reporting and/or recordkeeping requirements contained in this letter affect fewer* than ten respondents; therefore, OMS clearanc~ is not required under P.L.96-511 *
. If. you hav~ any q~estions*r~g~rding the ~nclosures, ple~~e contact the
. Dresden Unit 2 Integrated Assessment Project Manager, Greg Cwalina, at 301-492-8053.
Enclosures:
As stated
_cc w/enclosures: -
- See next page
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Sincerely, L1!!&.
Dennis M. Crutchfie ~. Chief Operating Reactors ranch No. 5 Division of *Licensing
-~
(d)
( e).
- High pressure coolant injection p1p1ng, fittings and valves - These components are made of Al06, Grade B carbon steel of various thicknes-ses.
The drain and condensate lines are exempt from testing due to thickness considerations.
The piping req~ires testing, but may be exempt from testing if the 1 i censee confirms that the 1 owe st service temperature, as defined in the ASME Code, is greater than 150°F.
Otherwise, the licensee must de~onstrate the adequacy of the fracture toughness for this component or demonstrate that the consequences of failure of the component are acceptable.
Condensate/feed 1 ~ter system piping - This piping is Al06, Gr~de B, with thickness ranging from 0.718" to 1.093".
This pi'ping requires* testing, but may be exempt from testing if the licensee confinns that the lowest service tempe.rature is greater than 150°F.
Otherwise, the licensee must Qemonstrate the adequacy of the fracture toughness for this compo-nent or demonstrate that the consequences of fracture of the component are acceptable.
Main steam system - This system is mad~ of Al06, Grade B carbon steel and is 1.031" thick.
This piping requires testing, but may be exempt from testing if the 1.i censee confirms that the 1 owe st service tempera-
- ture ; s greater than 150°F.
Otherwise, -the 1 icer:isee must demonstrate the adequacy of the fracture toughness for this component or demon-strate that the consequences of fracture of the component are.acceptable.
Since you have* not provided the necessary infonnation, the staff is unable to conclude that *adequate fracture toughness exists.
Further, you hav~ not supplied any infonnation regarding radiography requirements.
Therefore, the staff position identified in the draft IPSAR remains unchanged.
-The staff will require that the necessary infonnation be suppl i.ed in a revision to the updated FSAR within two years.
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DRESDEN 2, 3 *
- 5. n Question Pipe whip criteria that have been implemented in the design of the plant should be stated. A description of how these criteria have been used in the design of various engineered safety features should be included, with particular emphasis on ECCS piping and instrumentation systems located within the drywell. The description should also indicate how the effects of jet impingement forces on various safety feature. components have been.
- accounted for in the design.
Answer Pipe restraints to prevent pipe whip have been applied where deemed necessary to insure that:
- a.
containment integrity \\_Vill be maintained, b.. at least one core spray system, including instrumentation, will
- remain operable; and
- c.
at least one set of reactor pressure vessel.level instrumentation will remain operable.
It is felt that this criteria has been met by:
- a.
the application of pipe restraints to the recirculation loop,
- b.
physical separation of redundant ECCS piping and instrumentation, and
- c.
physical separation o.f level instrum,entation;
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DRESDEN 2, 3 5.22-2 Similar criteria has been satisfied under jet impingement forces through containment and penetration design, and the physical separation of ECCS comPonents.
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D 2. 3 I. A-1 QtiESTlO!'i
- I. A For certain of thPse itenis. the anal~*ses. research and de*:elopment, and design changes to pronde r£>solut10n have Ileen adequately described.
i-lowever, further technical infor-mat10n is nee_ded regarding the specific actions taken to provide adequate resolution of those items listed below:
- l.
Jet pump operation. monitoring. and system stability
- 2.
Pipe whipping and missile generation
- 3.
Pr£>ssure \\*e.ssel desil!n. with attention to b£>ll-mouthin~ and* vibratio11
- 4.
Independent r£>\\*iew of.vessel stress report
- 5.
- Periodic \\*essel inspection 6..
In-core flux monitoring instrumentation
- 7.
Consenatism in desi~i1 and fabrication of the primary system
- 8.
Core analvtical models
- 9.
Load confroi' ~*ith variable speed pumps
- 10.
Dr'sd£>n lock and dam failure A..\\'SWEFi.
Please not£> that items L -* 8. 9and 10 <trE' ::..11swercd i.-, Questions LB, I.C, I.D, I.E,
- rnd I. F, rrsp_ecti\\*ely.
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- 2.
Section :i. 2. 3 7 r;f '.hf? FSAR discusses th£> a;1a1yses which were rr:ade 0n potential
.mis~ile ~f'ner;).:ion :i1~idt' the cor.tai:<::11:-nt.
T~.e a~alyses* show that there a:re no sourcf'S 0: !"11issiles *.~*ithin the dryweil with the e:iergy pote11ti~!-~~~red to pene-trate the *c,rn.:a::o:*1e::; s!-.el:.
The recirculation lines h<tn: been provided with restraints to.limit the :notion of thE'se. Jinl's. The restraints are discussea*i~ Section 4. 3. 2 of the FSAR. The re-straints will limit an~* 1110tion of the recirc~lation iinesduring a PO!?tulated but highly im probablt* ru:1rr11la ti0n* pipe Tupture.
- The. examination of other piping system.s _ wii!_iin the dry well has led to the conciu-s ion that the main 5t'am lines and the feedwater lines contain suificient energy, should one *)f these lines sufier an instantaneous complete severance of the pipe in certain specific locations, that the broke~hne cou1d, possibly pen~t_r_ate th~ contain-ment shell. Therefore, studies have been made and tests have bee-n conducted to determine the failure mode of this piping; i.e.* tc)d_eJ~ri.lti~~J(-t_lie:-piping can sever completely and in a short enough time period to develop}he*..e.~~.i:gy~_gµ!.r~.<:i.
tn penetratP. the ront.1inment shell.
TC'sts h:iH* Ileen conducted as part of the AEC sponsored Reactor Primary Coolant.
Rupture Study which r1emonstrate that a relationship exists between the stze* or a
-~
cr:ick and the prnlJability that the crack will propagate rapidly. The applicability of this st*1dv tu the e\\'aluation of the problem of pipe rupture in the dryweJ.l has be-en discusst>d in detail in Oyster Creek. Docket 50-219, Amendment 34. The results of the tests indirat<' that for a crack of a size which gi\\'es a leakage of 5 gpm the probability *Jf rap.id propat;ration is 10- 6.
Thu~. a pipe which is cracked and for which the !Pak r:ite is approxim:itely 5 ~pm. thl*re is a probability of one in on~
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D 2, 3
- r. A-2 million times that the crack will propagate rapidly enough to result in complete sever~
ance of the pipe. If the crack were.of the size that resulted in a leakage of 15 gpm, there would he :i probability of rapid propagation of 10-4. This crack size and leakage rate is well within the leak detection capability provided for the drywell.
The leak detection capability in the drywell is discussed in Section 4. 3. 4 of the*
FSAR.
The conditions for critical (or. unstable) crack irrowth are based on the assumption that the cracks grow to critical size by mechanically or thermally induced cyclic loading or stress corrosion crackin~ or some other mechanism characterized by gradual crack growth, From the tests conducted and the rate of crack growth it can be conclud~::I that the in:iin steam lines and the feedwater lines will not suffer com-1 plete severance during the lifetime *of the p!ant. Therefore. restraints cannot be j~stified for tnese piping systems.
If an unde~ected f:iult could lead to.. rapid propagation or complete severance of any pipe in the drywell, it**woUld occur in a small diameter pipe, There are no failures which c:rn uccur in small diameter piµes which will lead to a penetration of the containment.:
Studies th:it h:ne been done for the Oyster Creek Plant to investigate the manner in' which pipe restraints would have to be designed to provide complete restraint capa-bility for* the ~team and fe~dwater llnes similar to the restraint capability that has been p*r.. wid~d in the Dresden 2, 3 plant design for the recirculation system piping have shown that it is impractical to design such restraints for the steam and feed-water piping because of mechanical and structural limitations from the point of view of anchoring these specific lines. In addition, the conceivable restraining devices that.would be installed would have to be installed*in* such a manner that the r~straints would prevent con.venient and ca.reful.inspection of the sensitive se~tions of these pipe lines without removing the mechanical restraint *devices at each inspection period.
- The Qresden Unit 2 and 3 main steam* and feedwater piping and the drywell struc-tural configurations are essentially similar to those of Oyster Creek and the prob-lems cf ~stalling restraints *are comparable to those encountered in the Oyster
. Creek investigation.
Therefore, it is our belief and recommendation that preventive mainteriance and
.regular inspection of sensitive pi~ runs is a more safe method to be followed in as-suring that large pipes in the reactor drywell will not fail, rather than assuming that very unlikely failures can occur and pro.viding ma.ssive restraining equipment that in themselves compromise the opportunity to perform maintenance and inspec-tion activitiPs.
In order to provide-the maximum assurance th;tt the emergency core cooling system pipini.:- :ind instrumentation will perform the required functions in the unlikely event of a pipe ru,>ture within the drywell. thC'se systems h:n*e i>l*en physically sep:irated
D 2. 3 I. A-3 and made redundant such that a postulated failure of any piping in the drywell will not disable or pre\\*ent proper operation of the emeq:ency core cooling system.
In addition. where µossible, ECCS pipinf; has been routed behind structural members i0r additional protection of the ECCS system.
- 3.
Be!l-mouthinl:< oi the reactor vessel is applicable only to vessels with breech clo-su 1*e, or a closure made by screwin~the reactor \\'essel head into the reactor vessel.
This ty~e oi closure is not used <m the Dresden 2, 3 reactor vessels therefore bell-mouthin!! is not applicable.
The Dresden* 2. 3 reactor vessels were designed and built in accordance. with the ASME Boiler and Pr~'ssure Vessel Code, Section 111.
The reactor vessel internals were desi>.!'tll'd with attPntion l>ein~ :,!;iven to \\'il>rations a11d the reactor vessel and internals were analy7.ed to determine their capability to withstand flow induced vi-ilrations.
Vibr:ition me:isurC'ments will be made. during the startup test period on the Drco-;~cn 2 rc:ictor \\*essel :ind internals to demonstrate the mechanical integrity of the system to,*ilJration motions. These measuren1ents are also designed to check the validity and :iccuracy of thC' analytical procedures used to calculate the vibration characteristics*of the system. The following points will be monitored for vibration:
- a.
Control rod !!Uide tubes.
- b.
ln-cnr(' ;::*1idP tulles.
- c.
fuel channds.
- d.
Core plate.
- e.
Shroud.
[.
Separators.
1!.
Recirculation loops.
- h.
Jet pumps.
- 4.
The vessel strf?ss report is being prepared by the Babc0ck and Wilcox Company*.
. An independent re,*iew of this vessel stress report is conducted by the General Electric Company on each section of the report as it. is received from Band W.
Upon completion of the analysis and the G. E. approval.* a certified report will be issued.
5 Detaj\\s of the periodic inspection program for the reactor vessel and primary sys-tem pipin~ a~e contained in the Technical Specifications, Section 4. 5,
- 6.
A detailed rPport ol the in-core flux monitorin~ instrumentation was submitted as topical report A PED-5706, December 1968.
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