ML19269E377
| ML19269E377 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 03/10/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1616, NUDOCS 7906280023 | |
| Download: ML19269E377 (59) | |
Text
'
...s.
e........:...
.,n.
rpt. 5//8/77 i
Acis - / 616
). ; rm;
.O 'l !.l!
DATE ISSUED:
3/10 /19 c
\\
v t
'I t
L U
MINlTTES OF 'IEE
- d//c/77 ACES SUBCOMMITTEE MEETING ON W4. H. ZIMMER NUCLEAR POVER STATION, UNIT 1 h%SHINGTON, IX' FEBRUARY 27, 1979 he ACRS Wm. H. Zimmer Nuclear Power Station, Unit 1 Subcommittee held a meeting on February 27,1979, at 1717 H Street, N.W., Washington, D.s'.
Notice of this meeting was published on 'Ibesday, January 2,1979, in the Federal Register, Volume 44, Number 1, and Friday, February 9,1979, Volume 44, Number 29; a copy is included as Attachment A.
A tentative presentation scuedule for the meeting is included as Attachment B.
A list of meeting attendees is included as Attachment C, and a list of background documents submitted to the Subcommittee and its consultants is included as Attachment D.
Dr. Richard Savio was the Designated Federal Employee for the meeting.
INTRODUCTORY STATEMENT BY THE CHAIRMAN ~
Mr. Bender, the Subcommittee Chairman, convened the meeting a' 10:00 a.m.,
introduced the other Subcommittee member and the consultants, and reviewed -
briefly the schedule for the meeting. He noted that the purpose of this
~
meeting was to continue the review of the application of the Cincinnati Gas & Electric Company (C3&E) for a license to operate the Wm. H. Zimmer Nuclear Power Station, Unit 1.
'Ihe Subc:xnmittee had received neither written comments nor requests for time to make oral statements from members.af the public.
EXECUTIVE SESSION A brief executive session was held.
Mr. Sender solicited comments from the Subcommittee and its consultants.
Indicating that the results of the recent Two Loop Test Apparatus (TLTA) tests conducted by the General Electric (GE) showed different pressure variations than the predicted ones, Dr. P.esset psked about the NRC Staff's opinion on the results of these tests.
He also indicated that he would like to know the relevance of these results to the (imme,r plant. 2169 001 7 9 0 6"2% 0 @
A
..., f..
w-e s
Wa. H. Zimmer, Unit 1 February 27, 1979 Dr. Catton provided some brief comments on the Applicant's written response, to some of the questions raised during the Subcommittee meeting on November 17, 1978. Some of the questions and the Applicant's written responses to those are included as Attachment E.
With regard to the response to his question on the fuel bundle lift potential, Dr. Catton commented that the arguments given (in one of the GE documents on this issue) as to why it is conservative to neglect the er ntrol rod friction in the analysis are not convincing. He called GE and discussed this issue and they promised to provide additional detailed infor-mation in the near futur9 In relation to the heat transfer coefficients used in the analysis of the drywell concrete temperature, Dr. Catton commented that these heat transfer coefficients were taken from McAdams' book on " Heat Transmission",'
and they are not the latest.
As an overall response to the Applicant's written response, Dr. Catton commented tb t although he still has some concerns in some areas, he feels that progress in being made towards resolution.
Dr. Zudans indicated that he would like to hear about the basis used for calculating the energy absorption capability of the pipe whip restraints.
PRESENTATION BY THE NRC STAFF Introduction - Mr. Peltier Mr. Peltier reviewed the changes made to the Safety Evaluation Report (SER) subsequent to the Subcommittee meeting on November 17, 1978 (Attachment F, Page 1).
Mr. Peltier noted that the issues identified in the SER have been categorized into three different groups:
2169 002 p-m
u __
Wn. H. Zimmer, Unit 1 February 27, 1979 Outstanding Issues - Includes those issues on which the NRC Staff
'has not completed their review so as to establish a final p>sition.
Confirmatory Items - Includes those items on which the review has been completed and the NRC Staff has established some positions.
he Applicant will implement these positions prior to the operating license issuance. %e NRC Staff is in the process of confirming the implementation of these positions.
Items of Disagreement - Includes those items on which there are some disagreements between the NRC Staff and the Applicant.
Mr. Peltier also indicated that, as a step toward improving the legal sufficiency of the SER, it has been modified in several areas (Attachment G) so as to outline clearly the bases for NRC Staff's conclusions reached on certain issues.
4 Outstanding Issues - Mr. Peltier Mr. Peltier noted that the SER identifies two outstanding issues:
(1) Mark II acceptance criteria, and (2) Emergency Core Cooling. However, since the issuance of the SER, the emergency core cooling issue has been considered re-solved. He discussed briefly the status of these two items:
Mark II acceptance criteria - The NRC Staff has reviewed the generic aspects of the Mark II containment system arx3 has reported its find-ings in NUREG-0487," Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria," dated October 1978.
Se NRC Staff was given to understand that the Applicant intends to take exception to two of the generic acceptance criteria. %e informa-tion provided by the Applicant to identify the extent to which he is committed to adopt the generic criteria is being reviewed by the NRC Staff. Upon completion of the review of all the itets pertinent to Mark II acceptance criteria, the NRC Staff will incitrie its find-ings in a Supplement to the SER (SSER).
2169 003
[
L Mn. H. Zimmer, Unit 1 February 27, 1979
,Emergnecy Core Cooling - Subsequent to the Subcommittee meeting on November 17, 1978, the NRC Staff has reviewed the results of the TLTA tests. We NRC Staff believes that sufficient margin exists in the present Zimmer emergency core cooling system calculations to ensure the adequacy of this system. 'Ihey also believe that the present emergency core cooling system evalua-tion model is in accordance with the requirements of Appendix K to 10 CFR Part 50.
However, if further analysis of the TLTA test results indicate that the present model is inadequate, GE will be required to revise the model accordingly so as to accommodate the new findings.
Load Combination Criterion on Zimer - Mr. Bosnak Mr. Bosnak noted that the Zimer plant has evaluated all the loads, both in the balance of plant system side and the Nuclear Steam Supply System (NSSS) side, by using the absolute sum method; there were only a few situations where the Zimer plant did not meet the absolute sum criteria, and appropriate measures are being taken to resolve these situations. We NRC Staff believes that the load combination criterion for the Zimmer Plant is resolved.
Mark II Lead Plant Criterion on Zimmer - Mr. C. Anderson Mr. Anderson reviewed briefly the Mark II Lead Plant Ioad Acceptance Criterion, indicating that the NRC Staff's findings on this issue are delineated in NUREG -0487.
He noted that 39 various load phenomena associated with pool dynamics were identified. After discussions with the Mark II Owners Group,14 load phenomena prepared by GE were found acceptable; 5 were identi-fled for specific plant review, and in 20 areas the NRC Staff developed criteria where they believed that the Mark II Owners Criteria were inadequate.
In relation to Zimmer, Mr. Anderson noted that among the 20 areas for which the NRC Staff has developed some specific criteria, Mark II Owners Group has adopted 8; another 6 areas have been resolved recently (Attachment H, Page 2); and 6 more areas are being resolved (Attachment H, Page 3).
2169 004
t Mn. H. Zimmer, Unit 1 February 27, 1979 Mr. Anderson discussed briefly the issues that are being resolved.
He noted that among the 6 issues that are being resolved, the most attention is given to two items: (1) Quencher air clearing loads; and, (2) LOCA Jet submerged drag loads.
In relation to the quencher air clearing loads issue, Mr. Anderson noted
.that the NRC Staff's criterion includes the magnitude of the ramshead loads; it also assumes that all of the bubbles released are in phase and at the same frequency (within a range from 4 to 11 hertz).
However, the Mark II Cwners Group feels that this criterion is unrealistically conservative. Rey have complied with the requirements of using the magnitude of ramshead loads; however, they have not complied with the requirement that all the bubbles be considered to be in phase and at the same frequency.
Wey have analyzed several cases for different valve settings and actuation sequences by using the magnitude of the ramshead loads, and also taking credit for the pipe lengths. W ey have also analyzed one case by assuming that the bubbles released are in phase and at the same frequency; however, in this case, instead of using the ramshead loads they used IWJ quencher type loads, the magnitude of which is less than the ramshead loads.
Mr. Anderson indicated that the analyses done for different valve settings by using ramshead loads and assuming different bubble frequencies gave higher containment response than the analysis which was done by using quencher type loads and assuming bubbles in phase and at the same frequency.
In response to a question from Dr. Plesset regarding a comparison between ramshead loads and quencher type loads, Mr. Anderson noted that the magni-tude of ramshead load will be about a factor of 4 higher than that of a quencher type load.
Mr. Anderson indicated that the NRC Staff has asked Zimmer and the Mark II Owners Group to document the methodology used in the analyses 2169 005
~,,
, m _.
Wm. H. Zimmer, Unit 1 February 27, 1979 t
and all the other informatien ecinent to this issue. W e NRC Staff will compare this with the KWU test data which they expect to get within two weeks, to determine the appropriateness of the methodology which uses quencher t,pe loads.
He also noted that Zimmer has committed to do some in-plant tests on this issue. He believes that this issue will be resolved in the near future.
With regard to the issue on LOCA je't subnerged drag loads, Mr. Anderson noted that the Mark II owners Group has proposed a new " Ring Vortex Model" which uses a different assumption than the model proposed by the NRC Staff. We NRC Staff's model assumes that the water blown out of the downcomer will be going out as a bullet (jet); however, the ring vortex model asstraes that the water from the downcomer will deform into a mushroom shape upon entrance into the pool. W e Mark II Owners Group believes that a substantial fraction of the kinetic energy of the jet is converted into a vortex ring. We Mark II Omers Group has recently shown rame films of scaled tests to the NRC.ataff to explain this model.
In _ response to a question from Mr. Bender as to why the Mark II owners Group prefers to use the " ring vortex model" instead of the " bullet (jet) type model", Mr. Anderson noted that the Mark II Owners Group believes that in a bullet type model the loads may be concentrated in a specific area which may affect some of the pipes and equipnent that may be directly below the downcomers; this may not be the case in the ring vortex model where the loads may be distributed over a larger area.
Dr. Catton commented that he is not convinced that a vortex will be formed at such an early stage.
Mr. Anderson noted that the film he had seen shows the formation of vortex at an early stage.
2169 006
...........T......
.......... ~..
(
Mn. H. Zimmer, Unit 1 February 27, 1979 The Subcommittee indicated that it would like to sce the film prior to the.
ACRS full Committee Meeting preferably on March 7,1979, and also like to get the description of the ring vortex model as soon as possible.
We NRC Staff and the Applicant indicated that they would provide the film and the description of the model prior to the 227th ACRS full Committee meetirs.
Mr. Anderson indicated that the ring vortex model is being reviewed by the NRC Staff and they expect to reach some conclusion in March 1979.
In relation to the issues on wCA/SRV (Safety Relief Valve) Air Bubble Drag load, Mr. Anderson noted that subject to the documentation of the re-ferences used by the Mark II owners Grcup, the NRC Staff will consider these issues resolved in the near future.
i Mr. Anderson reviewed briefly some of the propsed tests associated with the pol dynamic load confirmatory program (Attachment H, Page 4).
He 1
noted that the NRC Staff plans to discuss the detailed test plans with the Applicant within the next two weeks.
With regard to the issues on the pol dynamic loads, Mr. Anderson provided the following conclusions:
1.
%e NRC Staff believes that Zimmer has adopted a large majority of NRC Criteria.
2.
%e NRC Staff does not anticipate any problems in resolving the few remaining open items.
3.
S e resolution of the remaining open items will be included in a SSER in March 1979.
2169 007
/
g
(
Wtn. H. Zimmer, Unit 1 February 27, 1979
_4.
Updated information on the Mark II generic acceptance criteria will be included in a supplenent to NURS3-0487 in April 1979.
5.
%e NRC Staff believes that the confirmatory test program and Zimmer in-plant test program will pro-vide sufficient information to confirm the adequacy of the lead plant load acceptance criteria.
Status of the Evaluation of Reactor Cavity Support Analysis for Zimmer -
Mr. Kudrick Mr. Kudrick noted that the NRC Staff had evaluated the reactor cavity support analysis for Zimmer by using the existing one-dimensional code.
Recently they have received some information on the Best Estimate Analysis Containment (BEACON) code which is a two-dimensional code.
mey have compared the preliminary results of the BEACON code with the results of the existing one-dimensterral wde, and-they-have observed -~-----
significant difference; BEACON code results were approximately 50 percent of those computed by the one-dimensional code. Rey do not know the t
exact reason for this difference. However, they believe that there is no ' evidence that the results of the BEACON code will jeopardize the results of the one-dimensional code.
In response to a comment from Dr. Catton that he is not convinced that BEACON code will be the solution for the problems in this area, Mr. Kudrick noted that BEACON code represents advanced technology.
W ey have an on-going program to compare the various existing test results. hey realize that additional work is necessary to define the problem more clearly; however, they believe that based on the existing data they can make licensing decisions.
In response to a question from Dr. Catton, Mr. Kudrick noted that Los Alamos Laboratory will issue a status report on the comparison of the various test results in the near future, and that information would be made available to the Subcommittee.
2169 008
L --
c.;.
I t
Wn. H. Zimmer, Unit 1 February 27, 1979 In response to an earlier question asked by Dr. Plesset regarding the relevance of the recent TLTA test results to the Zimmer plant, Mr. Hodges noted that the existing ECCS model for the Zimmer plant meets the general requirements of Appendix K.
However, the MC Staff has asked General Electric to compare the existing ECCS model with the TLTA test data; should there be any significant discrepency that needs to be considered, General Electric will be required to make the appropriate changes in their ECCS model.
Items of Disagreement - Dewatering of Compacted Backfill - Mr. Peltier Mr. Peltier noted that there are some disagreements between the NRC Staff and the Applicant on this issue. Se NRC Staff requires that the water level in the encapsulated backfill shall be maintained at or below 457 feet above mean sea level measured at the backfill dewatering well. How-ever, the Applicant has not agreed with the NRC Staff on the maximum water level that should be permitted. He believes that the construction permit stage commitment specifie-3 480 feet above mean sea level and that 480 feet level will provide adequate protection against pore pressure in the com-pacted backfill. Se information provided by the Applicant to substantiate his, position is being reviewed by the NRC Staff and its consultants, and a resolution is expected in the near future.
Mr. Bender esked the Applicant to comment on the main reasons for the disagreements on this issue during his (Applicant) presentation so as to enable the Subcommittee to have a clear perspective of the issue.
Confirmatory Items - Mr. Peltier Mr. Peltier reviewed briefly the confirmatory items (Attachment F, Page 3). He noted that the NRC Staff has completed its review on these items and has established some positions. The Applicant is required to implement these positions prior to the issuance of the operating license. t e dis-cussions pertinent to some of the confirmatory items are as follows:
2169 009 Y
c...,
=
I Mn. H. Zimmer, Unit 1 February 27, 1979
(
4 Inservice Inspection Mr. Bender asked whether the requirements of the inservice inspection program for the Mark II containment are more difficult than those for the Mark I or Mark III containments.
Mr. Herman responded that the Mark II containment inservice inspection requirements are better than those for the Mark I because the equipnent is more accessible.
Recirculation Pump Trip Effects Mr. Wagner noted that the Applicant did rot include the effects of recirculation pump trip in the overpressuriza-tion analysis. However, as requested by the NRC Staff, the Applicant has subnitted a revised analysis to the NRC Staff, and that is being reviewed.
In response to a question from Dr. Catton, Mr. Peltier noted that the Zimmer Plant does not have automatic boron injection
,ystem. Mr. Tedesco indicated that the requirement of the
, automatic boron injection system for the Zimmer Plant will be decided along with the generic resolution of the Anticipated Transients Without Scram (ATd3) issue.
However, in view of the fact that the rulemaking process of the A'IWS issue may be time consuming, the NRC Staff will establish some interim re-quirements for the Zimmer Plant and will use these in the operating license process.
Automatic Actuation of Wetwell Sprays Mr. Peltier noted that the Applicant-is installing the necessary instrumentation required for the automatic actuation of the wetwell sprays (actuated 10 minutes after 2169 010 v-
m. H. Zimmer, Unit 1 February 27, 1979
(
the accident) in order ~ to increase the allowable bypass of the suppression pool during a small. break IDCA.
We NRC Staff is reviewing the necessary instrumentation diagrams to deter-mine the adequacy of the system to perform its intended func-tion.
Non-Safety Grade Equignent Mr. Wagner indicated that the NRC Staff has some concerns tout the use of non-safety grade equipnent to mitigate the consequences of some abnormal operationc1 transients.
We NRC Staff indicated that this is the first time this issue comes up in the licensing process. R ey consider this as a generic issue and they.are currently reviewing the adequacy of certain non-safety grade equipnent used to mitigate the con-sequences of some abnormal operational transients such as i
feedwater flow control failure, an a generic basis.
Mr. Wagner noted that the NRC Staff's position on this issue
, for Zimmer is that the equipnent relied upon to mitigate the most limiting transient (the excess feedwater event) be identi-fled in the plant technical specifications with regard to availability, setpoints, and surveillance testing. W e Appli-cant has been requested to subait his plan for implenenting the NRC Staff's requirements on this issue.
APPLICANT'S WRITTEN RESPONSE TO SOME OF THE QUESTIONS RAISED DURING THE NOVEMBER 17, 1978 SUBCOMMITTEE MEETING te Subcommittee. discussed briefly some of the questions raised during the Subcommittee meeting on November 17, 1978 and the Applicants written response to tho.Te questions (Attachment E).
(The comments pertinent to the adequacy of the Applicant's written response is included in these minutes under the Executive Session.)
2169 011
n-Nn. h. Zimmer, Unit 1 February 27, 1979 i
APPLICANTS RESPONSE 'IO NRC STAFF'S PRESENTATION - MR. FLYNN (CG&E)
Mr.'Flynn stated that the Applicant agrees with the status report provided by the NRC Staff on the emergency core cooling system.
W e Applicant providad some comments on the confirmatory items.
We discussion and commen ; on some of those items are provided below:
Toxic Chemicals (Route 52)
In response to a question from Mr. Bender as to whether the Applicant anticipates any problem in resolving the issue on toxic chemicals transported along Route 52, Mr. Flynn noted that there may be some problems in establishing the specific kind and quanity of chemicals transported along Route 52.
He noted that this issue has developed very recently.
Mr. Flynn noted that the Applicant is not able to perform an analysis of this problen due to lack of specific information on the nature of the toxic materials transported on Route 52.
- However, since the Applicant has some specific information on the toxic materials transported through the Chessie System Railroad, he performed an analysis assuming that the materials shipped by rail are in turn handled by trucks and transported along Route 52 to plants in the area of the Zimmer plant.
Information pertinent to this analysis is included in Attachment I.
In response to a question from Mr. Bender as to what the NRC Staff expects from the Applicant on this issue, Mr. Peltier roted that the Applicant should comply with the requirements delineated in Regulatory Guide 1.78, " Assumptions for Evalua-tion of Habitability of Nuclear Plant. Control Room During a Postulated Hazardous Chemical Release," which requires that the shipnent of toxic materials within 5 miles of the plant be analyzed if the frequency of shipnent is more than 10 times per year. W e NRC Staff learned that there are about 400 to 2169 012
i L.
i
(
Wn. H. Zimmer, Unit 1 February 27, 1979 600 vehicles pass by the Zimmer plant along Route 52 every day.
If the materials transported en Route 52 are only ammonia arxl chloride, there will not be any problem because the control room is well protected against the accidental release of these materials.
Ibwever, if some other toxic matarials are transported on Route 52, they should be analyzed as recuired by Regulatory Guide 1.78 and adequate design provisions, possibly a detector that would isolate the control room in the event of the release of some toxic materials other than ammonia and chloride, should be incorporated into Zimmer design.
In response to a question from Dr. Catton, Mr. Presky from Sargent & Lundy noted that most of the vehicles which carry toxic chemicals will use some sort of identification such as " Hazardous Materials".
Dr. Catton pointed out that if the vehicles carry some identi-fication symbol, it would be easy to get a good statistics of the
, nature of the materials transported along Route 52 by counting those specific vehicles for a very short priod of time.
Using the license plate numbers and other pertinent information, it would be easy to get the exact materials transported on Route 52.
Qualification of Ecuipment - Mr. Brinkman, CG&E Mr. Brinkman noted that the NRC Staff has reviewed the infor-mation on the qualification of the equipent on the balance of plant system side. 'Ihe NRC Staff has issued necessary criteria for evaluating the equipent qualification on the NSSS side.
'Ihe Applicant believes that he can begin the analysis of the NSSS side equipent prior to getting the operating license.
However, he be lieves that it will not be possible to complete the assessment of NSSS side equipent prior to the operating 2 } (19 O
L._
i Nta. H. Zimmer, Unit 1 February 27, 1979 t
i license issuance. Werefore, t.e believes that the NRC Staff should not make this as a strong requirenent that the assessment of the NSSS side equipnent should be completed prior to the operat-ing license issuance.
Use of Non-Safety Grade Ecuinnent Mr. Johnson from GE noted that.this is a new issue in the licensing process. GE does not agree with the NRC Staff that this is a generic issue. He believes that non-safety grade equipnent has been med for mitigation of abnormal transients. GE believes that such equipnent can be relied upon to perform its intended function.
In resp)nse to a question from Mr. Bender, Mr. Tedesco, NRC Staff, noted that the issue on the use of non-safety grade equipnent has been looked at by the NRC Staff in a generic sense. For the Zimmer Plant the feed-water transients are identified as the limiting ones. We
, NRC Staff will develop some technical specifications on this for the Zimmer Plant.
Indicating that it seems that the NRC Staff does not have any position in relation to resolving this issue, Mr. Bender asked whether the NRC Staff intends to delay the issuance of operating license for Zimr.ter until this issue is resolved.
Mr. Tedesco responded that the NRC Staff does not plan to held up the operating license for the Zimmer Plant because of this issue.
ue believes that appropriate steps are being taken te olve this issue prior to operating license issua 2169 014 9
L_
t t
Mn. H. Zimmer, Unit 1 February 27, 1979 l
INDUSTRIAL SECURITY (CLOSED) - MR. SCHUPT, CG&E Mr. Schott reviewed briefly the industrial security plan for the Zimmer Plant. He noted that the security plan for the Zimmec Plant is in con-formance with the requirements delineated in Regulatory Guide 1.17,
" Protection of Nuclear Power Plants Against Industrial Sabotage." A revised security plan intended to moet the requirements of 10 CFR Part 73.55 had been subnitted to the NRC Staff in May 1977, and it is being reviewed by the NRC Staff. 'Ihe results of the NRC Staff's evalua-tion will be reported in a SSER.
Mr. Schott indiccted tnat telephones, walkie-talkies, and appropriate
~
wireless rats will be used as communication channels. A security supervisor will be residing at the Zimmer Plant.
In the event of an emergency, the security supervisor will take appropriate actions.
In response to a question from Mr. Bender, Mr. Schott noted that the primary respansibility for the offsite security support lies with the Clermor.t County Sheriff. 03&E has radio communications with the Clemont County Sheriff's dispatcher.
In the event of an emergency, (E&E expects help from the Sheriff's Office within ten minutes.
In addition, G E can also get assistance from the Moscow County as well as from the Ohio State Patrol.
In response to another question from Mr. Bender regarding provisions to preclude the incidents caused by the insiders, Mr. Schott noted that all the Plant Staff will be thoroughly checked prior to their enployment.
In addition, medical evaluation will be performed on each individual by the medical officer to determine each persons mental stability.
Mr. Schott indicated that CG&E believes that the Zimmer security plan is as good as or better San any other security plans that have been licensed for other nuclear plants.
216?
015
A i
a l
(
Mn. H. Zimmer, Unit 1 Fe 1979 Applicant's Resconse on Mark II Containment Issues - Mr. Brinkman, CG&E Fr. Brinkman provided the followi.19 stat.uq report with regard to the Mark II containment issues:
1.
With regard to the issue on asymmetric pool swell load, he noted that the Zimmer Containment was evaluated in accordance with the NRC Staff's criteria on this issue.
He believes that the structure will take this load.
An evaluation report on this issue will ba submitted in March 1979 2.
Procedures for in-plant test programs and instrumentation i
requirements for in-plant tests are being developed.
Mr. Brinkman volunteered to send draft copies of the instrumentation requirements for in-plant tests to the Subcommittee and its consultants.
3.
'Ihe Applicant believes that using ring vortex model for evaluating the issue on LOCA jet sutraerged drag load will be a realistic approach.
Mr. Brinkman volunteered to subnit the details of the model and also volunteered to show a movie which ex-plains this pienomena.
'Ihe Subcommittee suggested that the Applicant try to send the movie on or before March 7,1979 so as to enable the ACRS full Committee to see this movie prior to the 227th ACRS meeting.
2169 016
t t
Wrn. H. Zimmer, Unit 1 February 27, 1979
',4. With regard to the item on LOCA/SRV air bubble drag load, Mr. Brinkman noted that they.are working with LaSalle and Shoreham to come up with a generic solution. E ey expect to provide a status report on this issue in March 1979.
DISCUSSION ON SOME SPECIFIC ITEfG Arrangements for Fire Fighting Personnel - Mr. Schott, CG&E Mr. Schott noted that the Applicant has made arrangements with the Washington Township Volunteer Fire Department; this is a volunteer group and has the prime responsibility' to respond in' th3 event of a fire at the Zimmer Plant.
In the event of an emergency, Washington Township Fire Department will send at least 5 personnel within 5 minutes to the Zimmer Plant. Washington Township Fire Department will be backed-up by the Richmond Fire Department (Vouteer Group), if necessary. Fichmond Fire Department personnel will be able to'come to the Zimmer Plant within 10 to 12 minutes.
Status of Applicant's Recruitment of Operating Personnel - Mr. Schott, CGCE Mr. Schott stated that the Applicant expects to have a maintenance engineer on board on March 5,1979. Eey already have a backup reactor engineer who has been on Zimmer Plant site for about four months. E ey are in the pro-cess of recruiting several other personnel as necessary.
Mr. Bender asked what contingency plan the Applicant has in the event of losing some key plant personnel.
Mr. Schott responded that in a situation like that they will designate some members of the plant technical staff to oct as back-up personnel. Rese back-up personnel wi-11 also go.through the same training process as the primary personnel.
In response to a question from Mr. Bender as to whether the Applicant has made any arrangements with other organizations to get assistance in the 2169 017
~
Ntn. H. Zimmer, Unit 1 February 27, 1979 event of an emergency, Mr. Schott noted that the y are developing back-up capability within the Zimmer Plant Staff; he believes that they will be able to develop sufficient capability within their Staff to handle the emergency situation.
Mr. Bender commented that the back-up personnel may not have adequate experience to handle the complex problems. He believes that in the event of losing some key personnel for one reason or another, it is advisable to make arrangements to have access to other resources to handle the situation.
In response to another question from Mr. Bender, Mr. Schott noted that if a back-up personnel is assigned to help a maintenance supervisor he will act as an assistant to the maintenance supervisor; he will not be involved in other tasks._ Mr Bender asked..whether the Applicant.has.made--
such a commitment to the'NRC Staff that there will be an assistant to the maintenance supervisor.
Mr. Schott noted that they have not maae such a commitment to the NRC Staff. However, they plan to do this.
Mr. Bender commented that the back-up program for Zimmer Plant seems inadequate.
He suggested that the Applicant should make arrangements with other organizatins to get assistance to handle an emergency situa-tion.
Mechanisms Established for Obtaining Information on and Utilizing Industry Experience - Mr. Schott, CG&E Mr. Schott noted that they have arrangements to receive information based on operating experience. %ey get such information from the NSSS Vendor, Inspection and Enforcement Branch of NRC and other utilities. Such informa-tion received will be discussed and incorprated into Zimmer design as appropriate.
2169 018
i
-t Wa. H. Zimmer, Unit 1 February 27, 1979 Foundation Settlement Measurements -- Mr. Herman, CG&E Mr. Herman noted that they have been measuring the foundation settlements at 22 different locations.
It is being measured at every three months.
'Ihey observed that the settlements at these points are uniform and within the limits as stated in the Final Safety Analysis Report (FSAR).
In response to a question from Dr. Zudans as to whether they notice any perturbations when the water level in the Ohio River went up, Mr. Herman noted that he is not aware of any such instance. He added that they did not make any special measurement during the high flood level period.
i Dr. Zudans suggested that it is advisable to measure the settlements l
dLring high flood level period.
Mr. Flynn noted that they will arrange for set.tlement measurements during the first week of March 1979, as they expect high flood level during that period.
In, response to some ques: ions from Mr. Bender, Mr. Heller, NRC Staff, noted that the NRC Staff is satisfied with the Zimmer Plant settlenent monitoring system. 'Ihe NRC Staff does not monitor the effects of the settlement at the Zimmer Plant because the Applicant has been measuring the settlement since 1975, and it was observed that all the settlements are within prediction.
Dr. Zudans asked.tethet the NRC Staff has reviewed the Applicants analysis of the settlemen effect on the structures.
Mr. Peltier it iicated that he will provide this information at a later date.
2169 019
e k
~
j Mn. H. Zimmer, Unit 1 February 27, 1979
(
Dew 5tering of the Compacted Backfill i
Mr. Herman from CG&E provided some bases to substantiate the Applicant's position on this issue that the water level in the compacted backfill can be maintained at 480 feet as agreed upon during the construction per-mit stage (Attachment.J).
Quality Assurance Program Indicating that a recent article appeared in the Cincinnati Post (Attachment K) gives the implication that the implementation of the quality assurance program at the Zimmer Plant seems to be inadequate, Mr. Bender asked why and how the NRC Staff believes' that the quality assurance program for the Zimmer Plant is adequate and satisfactory to them.
Mr. Conway from the NRC Staff responded that he believes that the quality assurance organization for the Zimmer Plant has the required independence and authority to implement effectively the quality assurance program require-ments. He stated that he is not aware of the article that appeared in the Cincinnati Post.
Mr. Vandel from Inspection and Enforcsunent Group of Region III, NRC, stated that they had performed several quality assurance inspections during the construction phase of the Zimmer Plant; they did not identify any peculiar problems related to quality assurance.
He added that the article appeared in the Cincinnati Post carried some misquotes and misleading statements.
Mr. Isa Yin from the Inspection and Enforcement Group of Region III, NRC, stated that he provided some information to the Cincinnati Post article.
His earlier inspection of the pipe hangers at the Zimmer Plant indicated that the workmanship, installation, and the record-keeping were not in accor-dance with the NRC Staff's requirements. Ibwever, since then some corrective actions have been taken by the Applicant, and he believes that now it is getting better. He noted that individual hanger support calculations are
~
2,t69 020
L -..
t I
Wm. H. Zimmer, Unit 1 February 2',
1979 being performed by Sargent & Lundy, and as soon as a major portion of the calculation is completed he plans to perform an in-depth inspection.
Mr. Bender sought some clarification of a statement in the Cincinnatti Post which states:
"Mr. Yin noted that about 50 of the plant's anti-shocking devices that cushion the pipes from earthquakes were of a make he considers unsound. Utility officials dispute this opinion."
Mr. Yin stated that this refers to the mechanical snobbers used at the Zimmer Plant. Se manufacturers (In'cernational Nuclear Safeguard Cor-poration) of these snubbers are no longer in business. Based on his experience, he believes that these snubbers may not perform the function it is Cesigned for. He also understood that some nuclear pwer plants such as Cooper and the Fast Flux Test Facility (FETF) were abandoning similar snubbers. During his recent visit to Korea, he also came to know that some of the snubbers (similar to the ones which are used at the Zimmer Plant) were completely frozen.
Mr., Bender sought some respnse from the Applicant on this issue.
Mr. Schwiers from CG&E indicated that the pipe hanger issue was under their surveillance prior to Mr. Yin's visit to the Zimmer Plant. tey did recognize that they had some erobleas. When Mr. Yin visited the Zinmer Plant, he identified some of the problems they already had. Wey are in the process of redesigning the hangers and they believe that they eventually will have an adequate pipe hanger system at the Zimmer Plant.
With regard to the. issue on thqi quality of.the snubbers -he noted that-r the Applicant does not have any strong evidence to indicate that these snubbers will perform its intended function. Therefore, they are in the p2 ocess of buying snubbers from a different manufacturer which are accept-able to the NRC Staff. Rey do not plan to discard the snubbers they have already bought; they intend to save them for the time being.
2169 021
u.
j I
Wm. H. Zimmer, Unit 1 February 27, 1979 t
In response to a question from Mr. Bender as to whether CE&E had discussed this issue with the p rsonnel at the Cooper Plant in view of the fact that they are also replacing these snubbers, Mr. Schwiers noted that they discussed this matter with the personnel at the Cooper Plant. Way found that the snubbers manufactured for the Cooper Plant were some what different; one of the differences is that the snubbers manufactured for the Zimmer Plant use all stainless steel. h ey are also checking to see whether the snubbers for the Zimmer Plant wre manufactured at the same time as those for the Cooper Plant.
Mr. Bender remarked that he believes that the NRC Staff will make sure that the problems identified on the pipe hanger system will be resolved to their satisfaction.
Huskie Cable Tray Allecations - Mr. Schwiers, CG&E Mr. Schwiers noted that the cable trays for-the-Zimmer Plant was supplied -
by the Huskie Company.
01e of the former employees of the Huskie Company made some allegations that the materials used by the Huskie Company in manufacturing the cP ~ e trays were inferior and rot in compliance with the specifications.
In addition, this former enployee stated that some of the welding on the cable trays was performed by unqualified wlders.
Consequently, CD&E conducted an independent review of the Huskie records.
The results of their review indicated that the materials used for the cable trays far exceeded the specification limits. m ey also conducted independent tests on some samples taken from the cable trays which verified that the materials used did exceed the specified requirenenti
- - by 30 to 35 percent.-- mey also-condteted some non-destructive examination:"~ ~
on the welds and found them satisfactory. As far as the qualification of the welders is concerned, there were tw welders, who had been with the Huskie Company since the mid 50's, who were not qualified for a very short period of time; however, these two welders were qualified shortly 2169 022
1 L_
hb. H. Zimmer, Unit 1 February 27, 1979 k
e afterwards. 03&E believes that this issue did not affect the quality of -
the cable tray fabrication.
Mr. Schwiers also indicated that a group identified as CARE and another group headed by Mr. r4ichael Bancroft of the Public Citizens Litigation Group recently discussed this Huskie cable tray allegations with the Region III Inspection and Enforcement Group of the NRC.
After completion of the scheduled presentations Mr. Bender solicited comments from the ACRS consultants.
Dr. Zudans indicated that he would like to get additional information on the energy absorption capability of the pipe whip restraints.
Mr. Bender asked the Applicant to provide the response to this issue in writing prior to the 227th ACRS full Committee meeting.
Mr. Bender also suggested that the NRC Staff and the Applicant should be prepared to discuss the status of the implementation of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident."
SUBCOMMITTEE REMARK 3
'Ihe Subcommittee indicated that it would recommend the operating license application of the CG&E to the ACRS full Committee for review during the 227th ACRS Meeting (March 8-10, 1979).
~
Mr. Bender suggested that the ACRS full-Committee may wish-to~ hear at least the following items in detail:
1.
Status of the Zimmer Plant (include construction status, fuel loading date, and other relevent information).
2.
Brief description of the plant by using a clear diagram.
2169 023-
..... ~.
i Mn. H. Zimmer, Unit 1 February 27, 1979 3.
Organizational Plans - How does the Applicant intend to establish the staffing.
4.
Transport of Toxic Chemicals Along Route 52.
5.
A1WS issue.
6.
Reactor flow control system.
7.
Status of Mark II containment issues.
Mr. Bender noted that should there be any other items identified for dis-cussion, the Applicant and' the NRC Staff will be notified immediately.
Mr. Bender thanked all the participants and adjourned the meeting at 3:00 p.m.
NOTE: For additional details, a complete transcript of the meeting is available in the NRC Public Document ~ Room,' 1717 H St.~,~ N.W.,
Washington, D.C. 20555, or from Ace-Federal Reporters, Inc.,
444 North Capital Street, N.W., Washington, D.C.
2169 024
)00R OR dy! L
' NOTICES
[7590-01-M]
" ** O l In addition, it may be necesr. ry for NUCLEAR REGULATORY the Sube mmittee to hold one or more closed sessions for the purpose of ex.
COMMd5 ION pbring matters *volving proprietary information.
, determined, in ac-Ad@ry Committe-en Reactor Sofegverds, Svinemailttee se the wuSea H. Zimmer W.
cordance with subsection 10(d) of deer Fewer $Mlea Pub. I.92-463, that, should such ses-sions be required. It is necessary to
{ close these sessions to protect propri-meting
(
The ACRS Subcommittee on the e cK4 William H. Zimmer Nuclear Power Further ' information. regarding Station will hold a meeting on Janu-topics to be discussed, whether the ary 17,1979, in Room 1046, 1717 a meeting ha.s been cancelled or resched.
Street, NW., Washington, DC 20555 to uled, the Chairman's ruling on re-review the application of the Cincin-quests for the opportunity to present nati Oas and Electric Company for a oral staternents and the time alletted
- permit to operate Unit 1 of this sta.
therefor can be obtained by a prepaid tion. Notice of this meeting was pub.
telephone call to the Designated FYd-lished on October 20, November 20, eral Dnployee for this meeting Dr, and December 20,1978 (43 FR 49080, R! chard P. Savio. (telephone 202/634-54147, and 59447, respectively).
3267) between 8:15 a.m. and 5:00 p.m.,
In accordance with the procedures EST.
w-w outlined in the FEDERAI, RrorsTot on Background information concerning October 4,1978,(43 FR 45926), oril or items to be considered at this meeting written statements may be presented can be found in documents on file and by members of the public, recordings available for public inspection at the will be permitted only during those NRC Public Document Room.1717 H portions of the meeting when a tran.
Street, NW., Washington, DC 20555 script is being kept, and questions ma?
and at the Clermont County Libmry, be asted only by members of the Sub.
Third and Broadway Streets, Batavia, comntittee, its consultants, and Staff.
Ohio 45103*
~
Persons desiring to make oral state.
ments should notify the Designated Dated: December 26.1978. <
' Itderal Dnployee as far in advance as pracilcable so that appropriate ar.
S4xun.J.Carut, * --
rangements can be made to allow the Secretary of the Commission.
necessary time during the meeting for tra Doc. 78-36424 Filed 12-29-78; 8 45 ami such statements.
I The agenda for subject meeting shall be as follows:
- : w:d.......f...
Wednesday, January 17, 1979--20:00 (7590 01.g}
%2.-- - a r a.m. until the conclusion of business.
~
- ~*'~
The Subcommittee may meet in Ex.
NUCLEAR REGULATORY.
ecutive Session, with any of its consul-COMMIS$10N tants who may be present, to explore and exchange their preliminary opin-ADVISORY COMMITTIE CN REACTOR SAFE-lons regarding matters which should GUARDS; $U8COMMITTit ON THE WitLIAM be considered during the meeting and K ZIMMER 1UCtEAR FOWit $fADON to formulate a report and recommen-dations to the full Committee.
Meet!$g teschedutid At the conclusion of the Executive The January 17,1979 meeting of the Session, the Subcommittee will hear ACRS Subcommittee on the William presentations by-and hold discussions H. Zimmer Nuclear Power Station has with representatives of the NRC Staff, been rescheduled to be held on Tues-the Cinelnnatt Oa.s and Electric Com.
day, February 27,1979, at 10:00 a.m. In pany, and their consultants, pertinent Room 1046, 1717 H Street, N.W.,
to this review. The Subcommittee may Washington, DC 20555. The. purpose then caucus to determine whether the of this meeting is to review the appl!.
matters identified in the initial session cation el the Cincinnat! Oas and Elec-have beer: adequately covered -and.
tric Com3any for a~ permit 16 operate whether the project is ready for Unit 1 of this station.
review by the full Committee.
All iterns pertah11ng to this Enecting remain the same as published on Jan-uary 2,1979 (44 FR 124).
FIDERAt f eor $TER, Vot. 44, NO.1-tut 3 DAY, JANUARY 2,1979 Dated: February 5,1979,
- Jome C. HoTu, Advisory Committee 2169 023
- ~~c o!"ccc LFR Doc. 79-4439 Filed 2.8-79; 8:45 am)
FEDttAl REGISTER, VOL. 44 NCL 21-FRIDAY, FistUARY 9,197p:
'IEm'ATIVE PRESENTATION SCHEDULE y
W4. H. %IMMER NUCLEAR POWER STATION, UNIT 1 SUBCOM 41TTEE MEETI!G WASHI?GTON, D.C.
FEBR'JARY 27, 1979 I.
Executive Session (ACRS - Open) 10:00 am - 10:15 am II.
Introduction (NRC Staff)
Include a brief summary of the significant changes made to the SSR subsequent to the Subcommitt e meeting on November 17, 1978 10:15 am - 10:30 am III. Status of Outstanding Issues (NRC Staff)
(Emphasis on Mark II Containment Issues) 10:30 am - 11:30 am 1.
Ioad combination criterion on Zimmer 2.
Mark II lead plant criterion on Zimmer 3.
Items need to be resolved prior to OL issuance.
IV.
Applicants Presentation 11:30 am - 12:30 pm 1.
Response to outstanding issues 2.
Discussion of Mark II containment criterion as it.. applies to Zimmer m
(Include any exceptions taken from the generic criteria and their significance) 3.
Industrial security (Closed)
LUNCH 12:30 m - 1:30 p V.
Items Identified at and since the November 17, 1978 Subcommittee Meeting (NRC Staff and 1:30 pm - 2:30 pm Applicant) 1.
NRC Staff implementation policy on Reg Guide 1.97 (NRC Staff - 15 mins.)
~
2.
Discussion of the Applicant's written response to questions raised at the November 17, 1978 Subcomnittee meeting (15 mins. - See.Atta:hment 1)_
'" '~
3.
Status report on allegations on non-specification work on cable trays (10 mins. - NRC Staff) 2169 026 A7TAcHMdNT b I
2-4.
General discussion to include presenta-tions on:
a) Arrangements for fire-fighting personnel (Applicant) b)
Status of Applicant's recruitment of operating personnel (Applicant) c) Mechanisms established for obtaining information on and utilizing industry experience (Applicant) d)
Method for making foundation settle-ment measurements (See transcript of Novec.ber 17, 1978 Subcommittee meet-ing, page 143) (Applicant) e)
Completion of the NRC reviews of Zimmer QA organization (NRC Staff) f)
Role of the Architect-Engineer in CA and OC organization as compared to other plants (NRC Staff)
VI.
Public Comments, if requested 2:30 pm - 2:50 pm VII.
Subcommittee Report 2:50 pm - 3:00 pn 2169 027 9
ATTACHMENT 1 ACRS SUBCOMMITTEE MEETING NOVEMBER 17, 1978 OUTSTANDING ITEMS REQUIRING RESPONSE FROM APPLICANT PRIOR TO THE JANUARY 17, 1979 SUBCOMMITTEE MEETING 1.
Transcript pages 13-15 Questions were raised regarding fuel bundle lift potential and the dynamic behavior of the downcomer tubes during blowdown.
These questions are addressed in more detail in the enclosed letter dated November 30, 1978, Catton to Savio, and on page 135 of the meeting transcript.
2.
Transcript pages 45 and 154 A question was raised regarding the film coefficient used in the drywell heat transfer analysis, regarding situations where af fferent types of conservatisms would be desirable.
3.
Transcript pages 156-160 The suppression pool time temperature history during blowdown was discussed. Questions were raised concerning relationship between pool temperature during the worst (from the standpoint of maximum pool temperature) LOCA and the test used to establish chugging loads. The enclosed letter dated November 30, 1978 also addresses this.
4.
Transcript page 164 Questions were raised regarding the differences between single downcomer tests and the conditions existing with the multiple downcomer.s in the actual suppression pool. The enclosed letter dated November 30, 1978 also addresses this.
2169 028 awnw8
ACRS SUBC0bMITTEE MEETING ON I?!. H. ZIMIER, UNIT 1 WASHINGION, D. C.
FEBRUARY 27, 1979 ATIINDANCE LIST ACRS NRC M Fender, Chaiman O utler M. Plesset, Member A. Hafiz I. Catton, Consultant F. Schauer Z. Zudans, Consultant J. Roe S. Duraiswamy, Staff C. Economos R. Savio, Staff
- J. Kudrick H. Brammer
- Designated Federal Egloyee R. Bosnak J. Kovacs ACE FEDERAL A. Bournia M. Meltzer R. Hemann C. Knowles B. Tumvlin J. Menning cr':#ER 500RE 6 CORBER T. Hazpster M. Wetterhahn R. Tedesco I. Yin STONE 6 IGBSTER ENG 00RP T. Vandel G. Dawe N. Wcgner J. Stolf GENERAL ELECIRIC CO I. Peltier W. Smith J. Convey R. Johnson L. Heller CINCINNATI GAS 6 ELECTRIC 00 SARGENT 6 LUNDY J. Heman R. Scheibel J. Seibert M. Jackson W. Schwiers C. Coulombe G. Ficke R. Givan E. Borgnann R. Cotta J. Flynn S. Rurka H. Brinkmann C. Krishnaswamy J. Schott R. Crawford R. Pruski A. Deguemendjian ATTACHMENT C 2169 029
-s LIST OF DOCUMENTS SUBMITTED TO THE SUBCOMMITTEE AND ITS CONSULTANTS Safety Evaluation Report by the Office of Nuclear Reactor Regulation, USNRC, in the Matter of Cincinnati Gas and Electric Company, Mn. H. Zimmer Nuclear Power Station Unit 1, NURD3-0528, Dated January 31, 1979.
2169 030 ATTACHMENT D
~
)
s a
f.;.%-G '
,jo p
g
-m.-
0'$ 'I
- , _,+L vj HE CINCINNATI GAS & ELECTRIC CO.1PANY ED '=
~ ~ '
CINCINN ATI.OMio 4 520e February 13, 1979 REcrnTo AW50?.T C0:,;.;t RFAcica sx :cug w,TTit Ox yg ac Dr. Richard Savio U.S. Nuclear Regulatory Cemission FE81 e 1979 Advisory Ccemittee on Reactor Safeguards lJj Washington, D.C. 20555 7g3p;gEggy,3g1, gig 1 %n)/,
y RE: OUTSTANDING QUESTIONS FROM THE NOVEMBER 17, 1978 WM H. ZIMMER ACRS SUBCOMMITTEE MEETING
Dear Dr. Savio:
This is in reply to your letter of December 19, 1978. Attach-ment A is the list of fcur outstanding questions from the November 17 Subcomittee Meeting which were not resolved at the November 28-30, 1978 fluid dynamics ACRS Subccmittee Meeting. Attachment B is correspondence from Dr. Ivan Catton to you concerning the Zirmer ACRS Subcomittee Meeting of November. While some of the questions in Attachment B overlap those of Attachment A, there are two questions which do not.
These are the Fuel Bundle Lift and Control Rod Tubing Location questions. Attachment C contains our responses to Attachment A and those two pestions from Attachment B.
As you requested, we are responding in writing in order to expedite the Subccmmittee Hearings.
It is our understanding that both the questions and the answers will become part of the record. Alsc, as you requested, these responses are being mailed directly to the Subcommittee Members and Consultants which you have specified.
Very truly yours, THE CINCINNATI GAS & ELECTRIC COMPANY Cf} &
yyyrsa By y
JAMES D. FLYNN, Manager Licensing and Environmental Affairs JDF: dew cc: With Enclosures Mr. Myer Bender Dr. Ivan Catton Mr. Harold Etherington Mr. I. A. Feltier (NRC)
Dr. Milton Plesset Dr. Zenons Zudans ggg ne Ti" f
Q f. % p o ? ? 'h :<'." p Q;3 R' g
. l ? i'. '.
- 3. j Arrecn u e 31T.
E
p,.
ACRS SUBCOMMITTEE MEETING NOVEMBER 17,1978
~
OUTSTANDING ITEMS REQUIRING RESPONSE FROM APPLICANT PRIOR TO THE JANUARY 17, 1979 7
SUBCOMMITTEE HEETING p.
.W i
- 1. Transcript pages 13-15
{
Questions were raised regarding fuel bundle lift potential and the dynamic behavior of the downcomer tubes during blowdown.
e These questions are addressed in more deta 1 in the enclosed h.
letter dated November 30, 1978, Catton to Savio, and on page 135 of the meeting transcript.
j t
2.
Transcript pages 45 and 154 A question was raised regarding the film coeffic.. ant used in the drywell heat transfer analysis, regarding situations where different types of conservatisms would be desirable.
N 3.
Transcript pages 156-160 i.
The suppression pool time temperature history during blowdown 11.-
tras discussed. Questions were raised concerning relationship p
between pool temperature during the worst (from the standpoint 7
of maximum pool temperature) LOCA and the test used to establish T
chugging loads. The enclosed letter dated November 30, 1978 k.h also addresses this.
C..
4.
Transcript page 164 t-Questions were raised regarding the differences between single N
v-downcomer tests and the conditions existing with the multiple i.-
downcomers in the actual suppression pool. The enclosed letter dated November 30, 1978 also addresses this.
.? -
/
LG
'A_.
2169 032 P
1
?
I E
,. b...
... a.;
. ~...,
~
ATTACHMENT "B"
Nove=ber 30, 1978 C5 /077
==
ABY130RT CtMITTtt oA l
E ictor s m cu us u.s. z u To:
Dr. R. savio P
From:
Ivan Catton AM h!
Subj ect:
Zinner Subco==ittee Meeting, 16-17 November 197h 8t LkULO M 0 4 3 6 D
l41 Copies To/M. Bender, ii. S Plesset J. Ebersole hmer.
I would like to add a few con =ents to several of the topics discussed
[' ','
at the Zinmer Subcoc=ittee Meeting. They are more generic than specific in L.
ature.
I' Reactor vessel Supports. One of the contributors to reactor vessel support loading is non-uniform pressure in the annulus around the vessel.
t.x The annulus pressures are calculated using a one-dimensional code when the M.
- M henomena is clearly two-dimensional. Large pressure variations exist and the flov being two-phase results in sonic conditions being predicted to exis~
l etween volu=>es specified for analysis.
This is accomodated by the analyst sing a Moody coefficient approach. It is not clear to me that the predicted
'.[-[,.
ressures are realistin or conservative.
k s.
Fuel Bundle Lift. Fuel bundle lift potential during the blowdown phase j
f a LOCA, a concern of Mr. Ebersole, is not fully addressed in any document lJ
!.l.
have access to.
As far as I can tell, the upward force on the fuel bundle s based on end of channel life friction'vith end of life bsing ~ defined as y
he time at which the friction between the control rod and the asse=bly J'
all is high enough to i=pede withdrawal of the control rod. With this f,.'
efinition the upward force a= bunts to 107 lb. The downward force, corrected
- l f
l or water buoyancy, is 300 lb. The net is then about 200 lb in the downward M
i i
v.e 2169 033 I
s irection. During blowdown, thei pressure in the bypass region falls faster j
han in the fuel bundle causing a decrease in the bundle to bundle gap.
It
~
s not inconceivable that a several fold increase in friction factor might'
.5
~
esult and the 200 lb margin might disappear. It is possible that I have f
ssed an i=portant aspect of the problem.
If so, I would like to reviev
'anc he documents clarifying how the fuel bundle lif t question is put to rest.
i Suppression Pool Downcocers and Lateral Loads. The downcomers are over i
hirty feet long and have no lateral restraints. The magnitude of the 7t:?
ateral loads depends strongly on the suppression pool te=perature and Nam-ovncomer mass flux. AcaresulttheloadsvarywiIhti=e.
If the pool emperature remains lov enough, large lateral loads do not occur and my oncerns are not well founded. To assess the potential for large lateral oads, the pool te=perature and downcomer mass flux time hister'.es are h
eeded. The pool te=perature must properly account for stratification.
F s'.
one consideration should also be given to the geometric arrangement of the i
ighty or so downcomers and possible interactions as the data base is pri-prily the 4T tests. The submergence is ten feet and restraints, if needed,
.mW ould be submerged eliminating concern about pool svell impact loads.
y w
Control Drive Tube Loestion. The control rod drive tubes are located
^;
b b, -
ery close to the recirculation line with half of the tubes being on each y
ide of the vessel. The tubes are routed so that if half are lost, due to N~
[4 pipe break, the remaining can still scram the reactor. It see=s to me
{,.
V hat the possible loss of half your control drive tubes is an important p-L onsideration and naybe they ought to be less exposed.
[,'
l.*.
o
/,
w.
p.:..
2169 034 P
E I
ATTACIIMENT "C"
Question No. 1 (Transcript Pages 13 to 15)
The downcomer piping is analyzed for static and dynamic loads.
The loading cases were obtained from the DFFR and are identified in detail in Subsection 4.3.1.1 of DAR, Amendment 1, October 1976.
The loading combinations are per Table 6-1 of the DFFR and are explained in Subsection 4.3.1.2 of DAR, Amendment 1.
The design limits are identified in subsection 4.3.1.3 and the analytical results are presented in Subsection 4.3.1.4 of DAR Amendment 1.
Tables 4.3-1 and 4.3-2 of DAR Amendment 1 are attached herewith showing maximum stresses in downcomers.
Since the stresses are within allowable limits, there is no need for providing a bracing system.
2169 035
ZPS-1-HARK II DAR AMENDMENT 1 OCTOBER 1976 TABLE 4.3-1 MAXIMUM STRESSES IN DOWNCOMER STRESS (PSI)
MAD (1)
VENT CONDITION (
CASE DESCRIPTION TYPE EMPTY FULL A
Dead Weight Static 436 436 B
LOCA Pressure ( }
Static 470 470 C
OBE Vibration Dynamic 7,428
)
OBE Pool Slosh Dynamic 523 D
SSE Vibration Dynamic 13,927 g)
SSE Pool Slosh Dynamic 981 E-1 LOCA Blowdown Dynamic 15,337 NA( I E-2 LOCA Vibration Dynamic 137 NA E-3 LOCA Fool Swell Dynamic 600 NA F
SRV Bubble Oscillations Dynamic NA 6,917 SRV Vibration Dynamic NA 4,262 (Resonant Sequential 1
Symmetric Discharge)
(1) As identified in Subsection 4.3.1.1 (2) For Analytical Convenience, LOCA pressure was assumed to act at all times.
(3) NA - not applicable (4) For Analytical Convenience, the larger " full" load case was used.
2169 036 E
ZPS-1-FIARK LI DAR AFIEND?!ENT 1 OCTOBER 1976
~
TABLE 4.3-2 DOWNCOMER PIPING ?!AXIt!Ut! CO>tBINED STRESSES
- EQUATION 9 ALLOWABLE STRESS STRESS LOAD C0h!BINATION CLASSIFICATION (psi)
(psi)
I Subsection 4.3.1.2.1 (a)
Upset 9,011 18,000 e
II Subsection 4.3.1.2.1 (b)
Emergency 11,957 27,000 y
III Subsection 4.3.1.2.1 (c)
Faulted 17,089 45,000 IV Subsection 4.3.1.2.2 (b)
Faulted 21,685 45,000
- The maximum stress always occurred at the drywell floor where the downcomer is anchored.
2169 037 E
7
a Question No. 2 A.
Transcript Page 45 In calculating the pressure respor._; in he drywell during a postulated loss-of-coolant 'ecidane-1 is assumed that there is no heat 'ransfer to the drywell walls.
As stated in Zimmer FSAR Subsection 6.2.'
3.2.1, the heat transfer coefficient betwera the dryv __ atmosphere and drywell liner was assumed to be zero in oruer to obtain the most conservative approach; whereby appropriate film coefficients were considered in the temperature analysis for drywell concrete under various boundary conditions.
B.
Transcript Pages 154 and 155 Regarding the analysis about the temperature of the concrete, the film coefficients for convection at drywell walls and floors are shown on Table 2-1.
The drywell and suppression pool temperature transient responses after LOCA were used as the boundary temperature changes.
A typical temperature transient response in drywell is shown on Taule 2-2.
Figure 1 shows the sketch of primary containment with the area of con-cern.
The temperature gradients for drywell walls and floors are shown on Figures 6, 10, 11, and 12.
4 2169 038 e
c
TABLE 2-1 Heat Transfer Coefficient for Convection #
- At 0.02 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after LOCA occurred Location Film' Medium h Durino Transient BTU / (HR) ('F) (FT) i l'
l-E,il,M,G Air (af ter 0.1 hr) 1.03 s
I Water 92.56 P
Water 85.44 J
Water & Air 35.60 K
Water 87.11 L
Water 43.55 E,H,M,G Condensed Steam 2000.00 e
2169 039 4 Reforc'nce:
" Heat Transmission", by William H. McAdams, Third Edition, McGraw-Hill, 1954
. t*
2
TABLE 2-2 Botandary Temperature Drywell Suppression Pool Time Temperature
. Temperature Hr.
@ Boundaries
@ Boundaries ey og 0.0000 135 9'O 0.00003 137 9'O 0.00005 155 90 90 0.00010 170
~
0.00050 230 92'-
0.00100 250 93 0.00200 265 95 0.00400 283 115 0.00600 285 125 0.00800 285 128 0.01000 285 130 0.02000 285 132 0.04000 275 135 0.06000 j 260 140
~ 230 144 0.08000 0.10000 205 146 0.:!0000 185 148 0.10000 180 155 0'.60000 383 159 0.80000 185 162 1.00000 187 165 172 2.00000 190 4.00000 200 195 6.00000 205 204
'8.00000 207 208 10.00000 209 211 20.00000 206 216 40.00000 190 200 60.00000 175 180 80.00000 170 165 100.00000 155 160 200.00000 135 152 2169 040 2-
P00R ORGIML Ficere 1.
The Sketch of Prirary Containment j,
with the Area of Concern s
- ;Y.
1
,m.~
'N DR'IWELL COIiTAIIC.iD;T WALL
,h 13
.1..
u..
.s.
CORE
]
. t.:
e
.,1. s. ".
C
. s....
.n
. "./..'
.-)l STEEL LINER PIATE Il
?.
\\i __
s
,g F
SACRincit.L (EICLccIcAL)
,r
$HIELD
....g i
.;**. p
..:p s'.~. ;
e
' e is l E.*
-..s
. Gm' i u.. _. -.
. -. _. -. n ':.
s 2; O 4
.~..
O:-
s
.b.. ',
e e
M PRopostO OPENgN3
..s.,
7.'.,. 4-[: ~~r...,..,,., :
+
t.-
.._e_
(,...g.<
f..1 k,...
p 8,..,
i 8
L DRYWF_d, FLOOR J
Q
.r..
>54
' ',, '/ SUPPRE3SIO:I CIU.;
^
l s ' ' ***L n:v r
Wg{
L_
_ _ ___6
- _._-. 7 LIIER-'
--~__dF ~ _'E.l
_..m._.J
..?.
_s=-
_ :I t
r
- t'
=
_:., PLATE 4<
- v..
/ -l g I) 4,-
IL
..,; R e
.t".!-c,".*",.
,. "t = v. ~.... % Q}
'J. -
K v.:........
_....,.:..~.9,
~.
t s
, v::.. :.. *. ::
.::...:.;.e.-.
y.
v S
d
'~
BASE MAT
(.,,
200 P00R OR 8 M.
~
m e.
I StCI.dy-5tato r.r.d Trancier.t Tc perature Distribution within Dryve11 Floor (f.rca f )
4 200 -
V.'i t,h C :..rs.a c oy H e clir.; Durio: Tr o..s ie n t j
[.
I
~ ~ ~ ~..). ;s op y
up op of
)
of 3 0 hr i
1 180
\\
,i I
s o
-Q
/
\\R
/
\\
30
/
d hr
,1
\\
160 i
hf o
e v
b e
\\
s
?
\\
10
=
y 140-i Q
h, h, \\'
h 2
4 j
c:
h N
\\
3o
\\
I A
o,o
/
12 t)
.0
/
h/
$ (p fU U!
N... __
o YO
.C
'E CONCRITE
-Ma r
i t
i I
i gan 0
10 20 30 3G 40 WA a *DU G.;,N..S S, :.;
.DQ gjgg Q42 g
l{
1i 3
i 1
250 L.
'"C3 e
O ji Figure 10.-
O N
Transtent Temperature Distribution with Condensation Assumption Dryvell Containment Wall (Area 1)
Q 200 I,
M C"3 1
3:=
n 1---
h 30 40 h
5.. i.
mi t.r
- i. )f,,"
f 00 sitaL LilitR S
Pt ATE I
2 a
CONCRtlE
.g g 50 sy 10 20 30 40 50 GO 70 72 0
E C 5, g4 e
WALL THICKNESS, X (IN)
~
ca
\\
300 l
1 250 i
'[
Figure 11..
T i
O j
Transient Temperateire Dis'ribution with e
l}
Condennntion Ac:tuqstion D:,;v.:-11 Contni:vvent k'all (Arca 2) y p'
c3
^ 200 r
O
.v F
I Z
p n)
P fd
+
D k
20 H' 15 o Il he
[1 30 3
M i
40 bl
\\
hr f3
[ k3 L..
N
> c.K N
[l*
l' jf
', th ' '.i.., l. 10,, ~
g s,
3 i
100 l
g/4 N
1 a
e j
- ,u N
i w.u
==
,gg'<
N e
1
~ Ng i
[
E 3
e q]c -
coucntsu-~
I t
I t
i I
8-C 50 0
10 20 30 40 50 54 A
E lh
/4
300R. OR Glk:1 t
c Figure 12.
1 l
Trancient Tc=perature Listribution dth l
Condcast. tion Accu ption F.or.cter Supportir.3 is2.1 (Area 3) l s
250
~
I m
4 Vs g
d D
200 --
P; e
i m
l~
2 s
s a
m e
8
- s
(.
o z
S 5 \\
.k 5 l! 5
^
150
~ 40,
3 d
s
,a 9-1 s s
.,,h r
.y.=
m.
M-,
~
~
2169 04.5 CONCRETE 100 8
I t
O 10 20 30 40 48 5 O
M WALL THICKNESS, X (IN) g.
'e QNESTIONS 3 AND 4 The main vent downcomer loading conditions are described in FSAR Section I.2.3.5 (page I.2.3-8 cf Appendix I), Section 3.3.1 (page 3.3-1), Section 4.3.1 (page 4.3-1) of the Wm.
H.
Zimmer Design Assessment Report.
The design lateral load of 8800 lbf applied at the tip of each downcomer represents an upper bound-load obtained from foreign licensee data reported in NEDO-21018 for maximum pool temperature and negligible air content.
Consideration of the grouping of downcomers in the specific Wm. H. Zimmer Plant geometry arrangement are also discussed in the Wm. H.
Zimmer Design Assessment Report in the sections referenced previously.
A comparison of the dynamic loading conditions measured on the downcomer in the GE-4T tests with the design load of 8800 lbf is presented in Section I.2.3.5 of Appendix I in the FSAR.
It is concluded in all cases that the design load as defined in the Wm.
H.
Zimmer Design Assessment Report is a conservative upper bound loading condition for design assessments.
Additional items raised by the ACRS regarding the pool tem-perature, pool temperature stratification, and downcomer ms-flux time-histories are addressed in the following para-graph.
Steam mass-flux time-histories,-pool temperatures, and_ air. con _...
tent are shown in Figures 1 through 3 for the Wm. H. Zimmer Station and also for a typical 4T test run.
The following data is shown in these Figures:
Figdre 1.
Flow Regime Map for Steam Injection:
T vs m-Recirculation Line Break for Zimmer Figure 2.
T vs m for 4T Typical Run and Recirculation Line Break for Zimmer Figure 3.
% Air vs m - 4T Typical Run and Recirculation Line Break for Zimmer Mass Flux Time Histories CONTEMPT /LT 026 was run to predict the necessary mass-flux time histories.
The recirculation blowdown transient was taken from the Wm.
H.
Zimmer FSAR.
Heat transfer on drywell..
walls was ignored in this calculation.
The pool temperature was maximized by requiring 100% liquid entrainment in the vent flow.
The air content is calculated as the ratio of air mass to steam mass (% of gaseous portion of blowdown).
The results compare well with the results given in the Wm.
H.
Zimmer FSAR.
2167 046 lb
.-.s.....--......
e
+
QUESTIONS 3 AND 4 (Cont'd)
Since detailed data on the 4T blowdown and pressure response were not available, blowdown rates were scaled from Figure 5-33 of NEDE-13468-P (Phase II & III 4T Test Report).
Results were comparad to Figure 5-28, the pressure response for a similar run.
The differential pressure (driving force for vent flow) compared well to test results.
The 4T transient is somewhat diffirent from the DBA because the 4T break was a vapor blowdown rather than a liquid blowdown and the break flow decayed more rapidly.
However, the 4T test did give a full range of steam flow rates at similar conditions to the Wm. H.
Zimmer Plant.
Although the pool was not heated as much by the 4T blowdown as the actual Mark II pools are expected to be, the runs done at the 1200 F initial pool temperature will bound the Wm. H.
Zimmer Plant condition.
These results are summarized in Figures 1 and 2.
The air content for the Wm. H. Zimmer Plant may be compared with the 4T test results shown in Figure 3.
It is seen that the range of air content is approximately the same in the test as in the Wm.
H.
Zimmer Plant.
Pool Temperature Stratification Pool temperature stratification' data is given in Figure 5-26 (NEDE-13468-P).
This figure shows a temperature profile (as elevation varies) in the pool at various times after the beginning of blowdown.
A large temperature gradient is seen after 20 seconds.
The gradient is much less at 40 seconds and later which is due to the excellent mixing promoted by chugging in the 4T test facility.
This is supported by traces which show the chuggin4 beginning at about 26 seconds.
However, even if one applies the maximum stratification observed in the 4T test to the entire Zimmer transient (clearly extremely conservative) local temperatures do not exceed 2050 F.
Assuming Lhat all steam condensation injected into the pool creates mixing and that chugging is a significant mechanism for mixing, it is clear -
0 F and 1500 F) 4T test will bound that the high temperature (120 all expected in-plant conditions.
2169 047 17
- 0 T vs $ Recirculation Line arcak for
-5 Ziusner L= 25.1CN..
t:ste: Flow P.cgite Boundaries Are Only Approximata S - 25.4 Ci4T Stcan Nass Flux Lb.3/Fe2-Sec O
g
- 10 15 23
?5 30
-i5 c
iC0 j
N 4
8-8 9
I Steam Escapes frcs the Pool (S) g y
\\
j ( vscillatory Interfaca 'a: Oc.ve.cc...ae Ex t t (1)
G"J a.
It /
05 p Jat (Ellipsoidal Spapa)(4a)'
20 "w
- i, L,
I"""
j ll Chugging (2c)
~3 Bubble Oscillation (3) e-
- j gjtargeLateraiLoad) 3 10 '.!
(Cone Shape. Oscillatory) j; y
'i:
g a
l '7
'l c..
.Z1rrr.er Recirculat1on Line Break g
t 1
t Chugging'(20)
"8 o
l<
d cS S
go,' '
(Large Vertic41 Load) i j
t 1
4 k
Ir.ternal Chugs (2a)
Jet (4c)-
~
.(Cone Shipa. Stable) m 1
- x
.t_
73 n
U' h
go "J6 800 QS jf0
,.f 75 N
Staam Mass Flux, Xg/,v.2.sec m
w
" Studies ol' Dynaraic loads in Pressure Suppression f
c RefereMe:
Conta".r.ent.." Catton :et. al., lJCLA. Presented at Sixth water O
Rocccor Safety Infore.ation Meating. November 6-9, 1978,
^
QGaic.'.ersburg.Md.
g
Recirculation Lino Srcak for Zience e
~
i a
l j
e l
}
~
c3 cs 17 3 s
so l
N g
i t
?
1 I
~
I S
'g' 80 Typical 4T Blowdown 2
l I
~_ "
Initial Pcol Temperatura 1200F
[
i
~~
.e
~
' ~
E
,,r a:::ll" '
Z CJl3,
toS g
i f
CD
,., Typical 47 61cwdawn 21.mer Rectrculation Line Braak 1
_ Initial Pool Temperature 700F I-l g
T' -
~
M CD CD
~
CL.
o J
yo s
s_
a g
30 tC 20 2:
30 '.
3[
2
~
5teara Mass flux. Lb.n/ft.3 c,
/
Figure 3
% Air vs m for 4T Typical Run and s
Recirculation Line Break for Zieae 5.G.
~~
i t
l
\\
'i t.0.'
T i-j<
j Typical 4T Slowdce a
i "llC3 s'-
O 0
3.o-M
- E C"3 2
"l2E' s
p P
8 1s--
t O
v
.u s
l.o WZir:n.er Recirculation Line Break N
,'g u
2g so 37 s
c' O
r lo o
u
=
ca 2
~
cn Steam P. ass Flux (Lbm/ft -sec) cp 20
CONTROL ROD DdIVE PIPING LOCATION During the November 17, 1978 km. H. Zimmer ACRS Subcommittee Meeting, the question was raised regarding the closeness of the Recirculation System piping to the Control Rod Drive (CRD)
Insert & Withdraw piping and given a break in the Recirculation piping what effects would rescit and would the reactor still achieve scram?
Design Considerations The CRD insert and withdraw lines (3/4" - SCH 80 piping) are routed such that half of the lines are on either side of the reactor vessel.
Appropriate design considerations were given to the effects of a postulated recire pipt rupture which would lead to pipe whip and/or jet impingement:
1.
Pipe Whip Restraints The Zimmer Plant Design considered the possibility of pipe whip and for the recirculation piping a pipe whip restraint system is provided.
The design provisions and criteria used to assure that the reactor and all essential equipment within primary containment are adequately protected against pipe whip are discussed in detail in FSAR Section 3.6.
High energy piping systems were analyzed for dynamic potential.
Both longitudinal and circumferential breaks were postulated.
Design criteria required pipe in any possible direction about a plastic hinge formed at the nearest pipe whip restraint cannot impact any structure system, or componeni important to safety.
FSAR Figures, 3.6-1 and 3.6-2 locate the pipe whip restraints as well as various breaks considered for the recirculation system.
FSAR Table 3.6-2 summarizes the results of the above analysis.
From this table, we confirm that no CRD piping is located close enough to the retirculation piping-to-be contacted during pipe whip.
2.
Jet Impingement The CRD piping is 3/4" - SCH 80 pipe which is routed in two 0
groups on either side of the reactor 180 a part.
Postulating a recirculation system pipe break the piping restraint system will restrict pipe move.nont (see FSAR Sections 3.6.4 and 3.6.5) but a resultant fluid jet force could possibly impinge on some of the CRD insert and withdraw lines.
(It is noted that the CRD piping is located on either side of the 2169 05i e
CONTROL ROD DRIVC TUBING LOCATION (Cont'd) recirculation pump and only some of the CRD piping would be affected by a break in the pump suction piping and conversely only some affected by a break in the discharge line.)
!!owever given that small pipes have high crush strength properties, the jet impact velocities will have an insignificant effect in compressing the CRD lines.
Additionally, there is no structure directly behind the impacted lines to support direct reaction loads.
Consequently, most of the impact energy imposed will either be transterred into kinetic energy or absorbed as strain energy in bending.
Therefore, compression of the piping is not of concern.
On the subject of crimping of small piping, experience tells us that a 3/4"
- SCH 80 pipe could be completely bent around a 24" pipe (this is not possible with a jet impingement force but helps to demonstrate a point) and would not be crimped c..osed (i.e.
minimal cross-sectional flattening).
A minimum of 3 gpm is required to accomplish scram and the piping would have to be completely sealed to prevent flow.
Thus, it is physically impossible to prevent scram by jet impingement or pipe whip against the CRD piping.
3.
CRD Piping Rupture The CRD design.is such that, if CRD piping should rupture, reactor pressure will act upon the drive piston causing rod insertion.
Neither jet impingement or pipe whip (becausa of restraints) could cause a pipe rupture howe.ver.
e 2169 052 E
270',
%$$f8R3i H*
s2 F13 S'e O '
V ss A5 CUMfEAENIA 8 EAK RHR
) ["oNc'TU$1N L
EA T
ANY ANGLE P AR ALLEL TO PIPE (
O BREAxio OwM p@
v keg g
8 gg 033 e wu. w. ziuu ca Nuc'_c Aa cwrca sT ATlON, UNIT I i
l e_
FIGURE 3.6-1 RECIRCULATION LOOP-A WITH' POSTULATED BREAKS
7:9 r27 f
2?9
/g 5:s ss
(
C)'
ri, 09 re N
a ss A%
^
.)L's's^T#252t!'L ANY ANGLE P ARALLEL TO PIPE (.
s7 O anE Ax io o '@
vM p@
J e,,
^
v t
~
sio 3,,
8 s,,
E 21M 054 NOTE: This fi;;ure is identical to 3.6-1 P00R ORGINAL
-. = r = c, = s :: :. m ~.1 FIGURE 3.6-2 RECIRCULATION LCOP-B WITH POSTULATED BREAKS
ZPS-1 TABLE 3.6-2 RESULTS OF DYNAMIC ANALYSIS ON RECIRCULATION SYSTEM PIPE RESTRAINT RESTRAINT DEFLECTION BREAK
- LOADING PEAK DYNAMIC DEFLECTION AT BREAK ID DIRECTION LOAD (kios)
(inches)
(feet)
RD! ARK S *
- S1 R
511 0.805 0.451 1/2-inch clearance 53 T
456 0.340 0.164 1-inch clearance S6 R
608 0.806 0.396 1/2-inch clearance S6 T
693 1.06 0.290 1-inch clearance 59 R
512 0.808 0.297 1/4-inch clearance 510 T
548 0.56 0.485 1-inch clearance D12 R
504 0.791 0.244 1/2-inch clearance _____ __
D3 R
135 0.239 0.705 1-inch clearance D3 T
482 0.395 0.272 1-inch clearance D4 T
270 0.151 0.423 Load on each restraint; l-inch clearance D(
T 362 0.181 0.613 1-inch clearance D8 R
257 0.448 0.422 Load on each restraint; l-inch cicarance D9 T
203 0.074 0.362 Load on each.
restraint; l-inch clearance F1 T
404 0.473 0.667 1-inch clearance
- See Figure 3.6-1.
2169 055
- Clearance is the physical gap between the pipe 0D and restraint cable or restraint f rame, respectively, as indicated by the loading direction.
b
.ges-1 TABLE 3.6-2 (Cont'd)
PIPE RESTRAINT RESTRAINT DEFLECTION BREAK
- LOADI::G PEAK DYNAMIC DEFLECTION AT BREAK
_, ID DIRECTION LOAD (kips (inches)
(feet)
REMARKS **
F4 T
157 0.058 0.312 1-inch clearance F6 R
181 0.355 0.498 1-inch clearance F7
?
162 0.059 0.991 Load on each res-traint; 1-inch clearance i
2169 056
- Sae Figurc 3.6-1.
- Clearance is the physical gap between the pipe OD and restraint cable or restraint f rame, respectively, as indicated by the loc 41ng direction.
E
FUEL BUNDLE LIFT A GE Licensing Topical Report,"BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings", NEDE-21175-2-P, demonstrates that
'BWR/6 fuel will not lift from its seat even in the event of combined SSE and LOCA and indicates that the effect of control rod forces may be conservatively omitted from the evaluation.
.This is generally applicable to Zimmer and other BWR/5 reactors also.
The tendency for the control rod to lift the fuel as a result of fuel channel bulge has been considered.
The amount of bulge required to prevent control rod settle should not be experienced since it would require channel re-use well beyond design exposures.
Operating reactor experience indicates that channels are replaced so as not to require control rod insertion if the control rod does not settle.
Even if the channel were used to the point where control rod settle did not occur, the drag on each of the four adjacent fuel assemblies would amount to only 42 pounds.
This force would be completely relieved within 0.030 inch of lift as the fuel nose piece moved on its seat since the calculated channel-blade interference to prevent control rod settling is 0.030 inches and the radial clearance actually provided for in the fuel support for the fuel nose piece is.030 inches (See attached detai-1-Al - If-the peak-LOCA pressure-differential-pressure were to occur at that instant, it is conservatively calculated that the resulting drag would be 75 pounds per fuel assembly.
This load would be resisted by friction in the fuel support wher,e it is reacted and the net effect would not be significant.
~
2169 057 27 a
c t
u TUEL ASSE13LY 4
LOWER TIE PLATE FUEL TO FUEL SUPPORT SEAT
~~
/
\\
's a
0.030" CLEARENCE
'ff " -
~
i
\\,
/
FUEL SUPPORT
}
Ii' -
c
,,/
TO GUID.I TUBE 5
'/'
SEAT i
1 E
L.._
l i
}
I I
l FUCL SUPPORT
(
l-j i
CORE PLATE-
/
- GUIDE TUBE
.s i
OCTAIL A P00RORIGNAL 2169 058
- 2. 6
CHANGES TO SER
'0UTSTANDING ISSUES
'CONFItiATORY ITEfiS
'ITEf1S OF DISAGREEf1ENT
'NEW ISSUES
' RESOLVED ISSUES
' BASES FOR CONCLUSIONS 2169 059
~
(VIEW GRAPH 1)
ATTAc HM ENT f
OUTSTANDING ISSUES
' MARK II ACCEPTANCE CRITERIA
-LOAD DEFINITION
-DESIGN ASSESSMENT
' EMERGENCY CORE COOLING
-TWO LOOP TEST APPAPATUS 2169 060 (VIEW GRAPH 2)
F
CONFIRMATORY ITEMS
'T0XIC CHEMICALS (ROUTE 52)
' QUALIFICATION OF EQUIPMENT
' TRANSIENT ANALYSIS (0DYN VS REDY)
' REACTOR FLOW CONTROL SYSTEM
' CONTROL R0D DRIVE TUBES
' INSERVICE INSPECTION
' RECIRCULATION PUMP TRIP EFFECTS
' EXEMPTIONS TO 10 CFR 50 APP, G, H, J
'LPCI FLOW DIVERSON EFFECTS
' PHYSICAL SEPARATION AND ELECTRICAL ISOLATION
' PROTECTION OF MOTOR / GENERATOR SETS 2169 061 (VIEW GRAPH 3)
F a
' AUTOMATIC ACTUATION OF WETWELL SPRAYS '
' SAFETY RELATED DISPLAY INSTRUt1ENTATION
'USE OF NON-SAFETY GRADE EQUIPMENT
' FIRE PROTECTION
' INDUSTRIAL SECURITY
' INITIAL TEST PROGRAM
' FINANCIAL (VIEW GRAPH 3A) 2169 062 P
JTEMS OF DISAGREEMENT
' DEWATERING 0F COMPACTED BACKFILL NEW ISSUES
' EMERGENCY CORE COOLING
'T0XIC CHEMICALS RESOLVED ISSUES
' REACTOR VESSEL SUPPORTS
' PLANT SUPPORT STAFFING 2169 063 (VIEW GRAPH f0 F r
~....,_...
BASES FOR CONCLUSIONS
' STANDARD REVIEW PLAN
' GENERIC TASK ACTIONS
' EXEMPTIONS TO 10 CFR 50
' PREVIOUS REVIEWS AND APPROVALS
' TOPICAL REPORT APPROVALS 2169 064 (VIEW GRAPH 5) g-
.=. m _.
. c,.., 2 m..........
....-.m.....
SUBCOMMITTEE NOVEMBER 17, 1978 CONCERNS
' CONTAINMENT CONCRETE TEMPERATURE
'DRYWELL FLOOR TEMPERATURE DISTRIBUTION
'DOWNCOMER DESIGN (LOADS, FLANGES, TEMPERATURE, MASSTFLUX)
' RECIRCULATION PIPE RUPTURE
' REACTOR VESSEL SUPPORTS AND ANNULUS PRESSURIZATION
' PIPE WHIP RESTRAINTS
' STRESS CORROSION ~(CONTROL R0D DRIVES)
' INSTRUMENTS TO FOLLOW COURSE OF ACCIDENT
' PUMP OVERSPEED (DECOUPLERS)
'WATERHAMMER
'PBF FUEL TESTS
' FUEL BUNDLE LIFT DURING LOCA 216')
065 (VIEW GRAPH 6)
F
..u,...a.....~....~....
~. w.. -;..,
'TWO ISOLATION VALVES OUTSIDE CONTAINMENT (%[,,[j/ c fee,:.,)
' LOSS OF 0FFSITE POWER (FLOOD)
' FOUNDATION SETTLEENT FIRE FIGHTING RESPONSIBILITY
'QA ORGANIZATION CliANGES
' INDUSTRIAL SECURITY
' AE - CONSTRUCTOR RELATIONSHIP
'ATWS STATUS (VIEW GRAPH 6A) 2169 066 P
ACRS SifBC0f01ITTEE MEETING, WIL'LIAM H. ZIMPER
~
~
'~
~
NUCLEAR POWER STATION, UNIT 1 2/27/79
SUMMARY
OF PRINCIPAL CATEGORIES OF ZIMMER SER REVISIONS MADE FR0" THE DECEMBER 1978 DRAFT FOLLOWING REVIEW WITH OELD 1.
Added conclusions and/or cited bases for the conclusions In this category, the deficiencies in the conclusions and bases in the SER draft were corrected by citing previously approved NRC documents that provided the appropriate basis for the staff conclu-sion (as an alternative to describing the detailed technical basis for a conclusion in the SER).
In this case, conclusions should cite either appropriate Regulations, Regulatory Guides, Standard Review Plans, approved Branch Technical Positions or industry codes.
cited in these references.
If the evaluation and conclusions relied on previous reviews, i.e. previous plants or topical reports, the SER conclusion should cite the specific plant SER where the matter had been reviewed and approved or, in the case of a referenced topical report, the NRC letter approving the referenced topical report.
The following are examples of the Zimmer SER revisions in this category:
Section 3.1.1 has been added to the SER to provide the bases for acceptance of the various industry codes and standards referenced in the SER. All of the codes and standards cited in the report and enclosed bibliography, have been incorporated in our Standard Review Plans and thereby approved as accepted industry practice. (Section 3.1.1)
' Recirculation pump overspeed during a LOCA (3.5.1)
' Tornado missile protection (3.5.3)
' Barrier design procedures (3.5.5)
- Prote' tion against pipe whip (3.6) c
- Seismic input criteria (3.7.1)
' Seismic instrumentation (3.7.3)
- Containment design (3.8.1)
- Seismic qualification of Category I I&E equipment (3.10)
' Fuel densification; fission gas release calculations; instrument tube vibration wear effect on channel box corners; pellet / clad interaction; and post-irradiated fuel examination (4.2)
' Nuclear design description (4.3.2) 2169 067 RTTacHMGNT &
- GETAB review (4. 4.1 )
- Reactor internals materials corrosion (4.5.2)
- 0verpressurization protection (5.2.2)
" Corrosion of reactor coolant pressure boundary materials (5.2.3)
- Calculation of drywell floor upward pressure using CONTEMPT LT computer code; and subcompartment pressure analysis using RELAP-4 and COMPARE computer codes
( 6. 3.1 ).
" Residual heat removal cystem NPSH during containment cooling (6.2.2)
' Containment inerting evaluation
('6. 2. 5 )
- Criteria used for review of safety-related instrumentation (7.1.2) and control. systems (7.1.3)
" Surveillance testing _frecuency of engineered safety features system; feedwater isolation valve control (7.3.3)
- Reactor water levet instrumentation -
(7.5.3)
- Leakage detection system; neutron monitoring system (7.6.3)~-
Undervoltage or underfrequency protection of. Class IE equipment (8.2.3)
Spent fuel cooling system capacity criteria (9.1.3)
' Control room ventilation and air conditioning system (9. 4.1 )
- ECCS equipment areas ventilation systems (9.4.3)
Diesel generator fuel oil storage and transfer system (9.6.1 )
- Diesel generator auxiliary systems (9.6.2)
- Emergency planning (13.3)
' Radiological consequences of turbine trip or generator load rejection without credit for man-seismic equipment
-(15.2) 2169 06'8
. 2.
Clarification and/or supplement to evaluation In this cate r
to either, (gory, clarifying revisions we' e made to the SER drafta) define the in its evaluation or (b) supplement and/or correct the SER text.
The following are examples of Zimmer SER revisions in this category corresponding to (a) or (b) noted above.
In some cases, we changed our conclusion and determined that there was still an outstanding matter to be confimed;* in other cases, we determined that a previous concern was not a safety matter affecting plant licensing.**
(a)'Adequacyofthepost-accidentradiationdoseof 2.6 x 10 rads (3.11.5)*"
(a) 'BWR channel box deflection or distortion from irradiation growth; fuel rod bow (4.2)**
(b) 'Complian:e with Section 50.55a of 10 CFR Part 50, " Codes and Staidards" (5.2.1)
(b) ' Reactor coolant pressure boundary preservice inspections and testing program relief from 50.55a(g)(2)
(5.2.4)*
(b)
Fracture toughness evaluation and material surveillance
( 5. 3.1 )
_ program.non-compliance,with,AppendiceseG & H,to 10aCFR
-(5.3.2) -
Part 50 (5.3.3)
(a)
- Containment leakage-rate acceptability basis (6.2.1)
(a)
- Steam bypass capability between drywell and suppression chamber (6.2.1 )
(b) Mark II containment design assessment and closure reports; pool dynamic load acceptance criteria (6.2.1)
(b) ' Transportation of toxic materials on U. S. Route 52 (6.4.2)*
(b) ' Separation of reactor protection system power sources (7.2.3)*
(a) " Main steam line isolation valve closing time limits basis (7.3.3)
(b) " Reactor manual control systems; feedwater control system; pressure regulator and turbine controls (7.7.3)*
(b)
- Criteria used in evaluation of electric power system (8.1.2)
(a) ' Grid stability ~
('8. 2. 3 )
(a)'Groundingevaluation (8.3.3)
(b) ' Fuel storage rack spacing (9.1.1) 2169 069 En
(b) " Overhead crane evaluation (9.'.;
(b)
- Compressed air system evaluation (9.2.-
(b)
- Seismic and quality group classification criteria of the (11. :. " )
liquid radwaste. equipment;.the-process offgas system;_
(11.2.-)
and the solid radwaste system (11.1.2)
(ll.*)
(b)
- Plant operating organization structure (13.1)
(b) ' Emergency planning conduct of evacuations (13.3).
(b)
- Initial test program (14.1 )
(b) " Criteria used for evaluation of radiological consequences of control rod drop accident (15. 3..' )
(b)
- Plant Q/A organization structure
- (17.1 )
(b) *Effect of delay of one year of plant operation on financial evaluation (20.1)
(b) "ACRS generic concern, Group II E on soil-structure interactions (APP B-4)
~ 3.
Bases of acceptability of licensing pending resolut#;n of a generic
~
~
problem In this category, there were deficiencies in the staff evaluation of some generic problems; that is, an inadequate basis was provided to support'a conclusion that the plant could operate pending resolu-tion the problem.
The basis of proceeding with the issuance of a license should either be explicitly defined in the SER or related to the generic activities listed in NUREG-0371 and NUREG-0471, which in turn provide the bases for licensing of plants pending resolution of the generic problem.
The following are examples of Zimmer SER revisions in this catege y:
' Fuel assembly response to seismic and LOCA loads (4.2
' Thermal-hydraulic stability (4.4.1
" Control rod ' tube cracks (4.5.1 2169 070 e
. 4.
Changes to " Confirmatory Items" listing in Section 1.9 cf the SER.
These are items for which we have reached positions which the applicant will implement, but need documentation from the applicant re;arding the implementation of these positions.
ITEMS ADDED SER Sections 2.2, 6.4.2 - We have not been able to define the hazard, if an/, to the control room operators resulting from the transport of toxic chemicals along U..S. Route 52. past the _Zimmer site.
We are pursuing the needed infonnation.
SER Sections 4.4.0, 4.4.2, 15.1 - We are evaluating a riew calculational basis (ODYN Code) for the load rejection without turbine bypass.
We exrect to complete this evaluation prior to Zimmer licensing.
SER Section 4.5.1 - We are investigating evidence of cracking in control rod drive tubes and have requested additional inf.ormation on this matter from the applicant.
SER Se: tion 5.2.3 - The applicant will also provide inf'nnation regarding implementation of the positions stated in NUREG-0313 rsiated to stainless steel cracking.
SER Sections 5.3.1, 4.3.2, 6~.2.6~- We~ are reviewing information to determine whether or not we may grant certain exemptions to 10 CFR, Part 50 Appendices G, H and J.
SER Sections 7.1.3, 7.4.4, 7.6.3, 7.7.3 - We will also resolve the issues regarding range, setpoint drift, response time testing and test frequencies prior to issuance of the technical specificai. ions with the operating license.
SER Sections 7.3.3, 7.5.3_- We have not completed our review of instru-mentation required for the automatic actuation of wetwell sprays and monitoring of essential drywell and wetwell parameters in the event of a small break in the primary system.
Section 14.1 - We will complete our evaluation of changes to the startup test program and the preoperational tests of the essential direct current systems in a supplement to this report.
Section 20.4*- We will update our financial conclusions hen updated information is available from the applicant.
2169 071 61
' ITE"S RESOLED (DELETED')
Section 14.1.3 - Applicant's program fcr providing accep at:le plant and support staffing for the preoperational and startup ss or grams.
5.
Added Section 1.10 - Items of Disagreement Between the S aff and Applicant During the review a disagreement regarding the appropriate level for dewatering the compacted backfill under Category I structures developed between the applicant ar.d us.
We will resolve this matter with assistance from our consultant (see subsecticn 2.5.3 of this report).
2169 072
ZIFTER MARKII LEAD PUWT LMD CRITERION OVERVIEW 39 LOAD OR PHENCMENA SPECIFICATIONS P
14 ORIGINAL MK 11 CRITERIA ACCEFTABLE 39 5 Puwr ik1IQUE REVIEW 20 NRC DEVELOPED CRITERIA r
8 ADOPTED BY 'lK II 0,G.
20
< 6 RECENTLY RESOLVED 6llNDERREVIEW (2 SIGNIFICANT AREAS) 216c) 073 ATTHCHN6nT H
ZIMER ik II LOADS RECENTLY RESOLVED Imo/ PHENOMENON RESOLUTION 1.
SUBMERGED BOUNDARY
- EVALUATION OF ZIMER DURINGVENTCLEARING CONTAINMENT
- LETTER REPORT - MARCH
- 2. SMLL STRUCTURE ADOPTED 11RC IMPACT
- l. DAD CRITERIA
- 3. ASYMETRIC
- EVALUATION OF ZIMER CONTAINMENT POOL Swii
-LETTERREPORT-MARCH ll.
"T"QUENCHERARMLOADS USE N METHOD FOR FOUR ARM QUENCHER 5.
'T" OUENCHER USE DFFR METHOD FOR FOUR ARM TIE-DOWNLOADS QUEfGER 6.
"T" QUENCHER ZONE OF INFLUENCE CYLINDRICALZONEOFINFLUENCE 2169 074 H
...... ~ ~..
ZimER MK II LmDS UNDER REVIEW LOAD / PHENOMENA RESOLUTION 1.
QUENCHER AIR
- 6 LOAD CASES MEET INTENT OF NRC CLEARINGLOADS CRITERIA 0F SRV LOAD f%GNITuDE, FREQUENCY AND PHASING, CONFIRM WITH M TEST IATA, MARCH 1979.
-IN-PLANTCONFIRi%TORYTEST PLANNED.
2.
LOCA JET SUBMERGED DRAG
- llEW RING-VORTEX MODEL
-PRELIMINARYSTAFFREVIEW ikRCH1979 3-5, LOCA/SRV AIR BUBBLE DRAG
-ACCELERATIONDRAGCOEFFICIENTS
- EQUIV, VELOCITY IN A UNIFORM FLOW FIELD,
-IhTERFERENCEEFFECTSCLOSELY SPA:ED STRUCTURES
- GENERIC REPORT lhRCH 1979
- 6. CHUGGING FSI
- GENERIC RESPONSE TD lEC QUESTIONS M CH 1979
- IP CmFIRt% TION 2169075
POOL DYNAMIC'lMD CONFIRI% TORY PROGRAM
- ZImER IN-PLANT TESTS
- TEST PLAN [hRCH 1979.
- Exm0ED 4I TESTS
- CONDENSATION OSCILLATIONS
-PROTOTYPICALVENTLENGm
- REFORT 40 1980 -
- ibt GRi II - TEST PROGRAM
- CONDENSATION OSCILLATIONS v'NT LATERAL lmDS
- PROTOTYPICAL OF A SPECIFIC PLANT
- DATA, l'hY 1980 2169 076 H
CONCLUSIONS ZineR POOL DYNAMIC LOADS
- ZINER Ao0PTED l.ARGE Ih)0RITY OF 11RC CRITERIA
- ANTICIPATE lb PROBLEMS IN RESOLVING FEW OPEN ITEMS
- ZinER SER SUPPLEMENT MCH 1979
- GENERIC SUPPLEMENT TO IUREG 0487 APRIL 1979
- Ih Il CONFIRt% TORY PROGRAM AND ZineR IN-PLANT TESTS CONFIRM LEAD PuwT LOADS 2169 077
~
H
TOXIC CHEMICALS A review of the hazardous materials transported on the Chessie System Railroad has been performed.
Since no data exists for truck shipments, it is assumed that the chemicals shipped by rail are in turn handled by trucks.
Assuming that the chemicals are shipped by truck using the maximum size of 50,000 lbs, those toxic chemicals that are shipped more than 10 times per year (As per R. G.
1.78), are listed below:
Chlorine Carbon Dioxide Vinyl Chleride Xylene Aniline Acrylonitrile Anhydrous Ammonia Butadiene Butene Sulfuric Acid The control room has been equipped with redundant detectors for ammonia and-chlorine.
Since the toxicity limits of carbon d'ioxide, vinyl chloride, xylene, butadiene and butene are very high, they do not pose any hazard to the control room habitability.
Sulfuric acid is not a hazard to the plant as discussed in response.o question Q312.25 for rail 2169 078 ATTACHNSW Y 1
TOXIC CHEMICALS PAGE 2 shipments of toxic chemicals.
The only chemicals with relatively low toxicity limits are aniline and acrylonitrile.
A review of the industries within 10 miles of the plant has confirmed that these chemicals are not used or produced by the industries in the area.
Therefore, it is highly im-P probable that these chemicals are shipped frequently by trucks in the vicinity of the plant.
2169 079
_I 2.
t TABLE 2.1-4
)
DtDUSTRIES WITRIN 10 MII.ES OF THE SITE DISTANCE AND DIRECTIQ1f ICGIX CF
^ ~
EAZARDOUS MODE VF DfDUSTRT FROM THE SITE D9LOTIES FRODUCTS MATERIALS TRAN3 Pot?
t Moscow, Chio 0.3 mile ESE F & 3 Machina shop (1)*
10 Job shop aschining Elas Mfg. Co., Inc. (2) 25 surgical appliances and supplias
}
Carntown Ey.
i 2 milas S5W Elack River Mining Co. \\
150 Metallurgical, 40E Ritrostarch, Truck (3) limestona base powder with ammonia nitrate I
primer (underground -
6000 lb surface. 7300 lb)
Pt. Pleasant, Chio 2.5 miles NE Pickney Irewar Co. (4) 7 Crushed limestone 23,000-lb Truck magazina T3T Footar Ky.
4.7 miles SSE Clyde 5. Farkar taady--
4 taady-sized concreta.
Mix concreta (5) decorative patio blocks Ubeat Stone Caramics, 6
Pyrometer protection Inc. (6) tubes, the nocouple insulators, ceramic ferrules New sfeh=and, Chio 6.8 miles ENE Classena J & E Co. (7) 185 Woolen, worsted synthetic yarna - --
Sargent Tool tr Mfg.
1 steel parts, plastic Co. (8) injection meAas Talicity Chin 7.7 miles E Special Fallet Ca. (9) 10 Fallets Butler Ey, 9.3 miles 59 Butler Products (10) 78 Prop legs for Pressurized oxygen Truck trailers tanks Acetylene Truck Griffin Industrias, Isa. (11) 52 tendering tallow grease, meat meal, hides Eardy Brothers Mills (12) 1 Feed Jay Gee, Inc. (13) 10 Machine shop, estal fabrication
. Jessie Tavter Mfg. Co.
1 Screw machina (14) products W. C. Beckjord Gener-9.5 miles NNW
-236 Electric Fover ating Station (15)
(Chio River Mila 453)
Alexandria, Ky.
10.6 miles IIU Alexandria Plant of 8
Esady-eized concrete - -
1111 top Concreta Corp. (16)
Floet-Hi-Balsa Float 10 Fishing floats, c.orks.
Co. (17) bobbers tais concrete Co. (18) 7 Concrete septic tanks, pipe, ready-sized
~~~
216')
080 SOURCES: 1973 Kancucky Directory of Manufacturers. Ey. Dept of Coc.eres, 1974 Directory of Ohio Manufacturers Chio Chamber of Commerce.
j
- Nunhers in parentheses relate to locations shown on Figure 2 2-1
p..
I" 136
^
comment?
l We are prepared to go into it briefly.
.r MR. BENDER:
Yes, if you can do it briefly.
~
MR. HERMAN John Herman.
de feel that the PSAR states that, should.the water level beneath the F
,b V
critical buildings, exceed elevation 480, it states a pump will be installed.
a e
Now, the -- at elevation 480, if the water ever w
got to that elevation, 36 percent of the surface. area at that I-i i
elevation would be unsaturated, therefore, the water at i
[-
this elevation would be -- still have a capacity to relieve i
r the excess fore pressure due to any liquefaction of the surrounding material.
i
- Now, in March of 1978, we had a flood up to elevation. --...
1 493, and a water level that was calculated would rise inside the clay blanket at the rate of six-hundredths o.
a f oot per day.
t If you would postulate the design flood of 508.5, i
i that same water level increase would be nine-hundredths L.
of a f oo t pe r da y.
There is no flood on the Ohio River that ~would last of a sufficient duration to get the water up to elevation i
480.
7 So it's cur point that we f eel that the cump is not necessary due that the water evaluation inside the clay
(.
2169 081.
t F
N.
p.
P.
137 g -.
blanket would never get to elevation 480.
e r
In December of 1978, we had another flood.
The flood got to elevation 497 and the water inside the clay blanket ose to -- rose six inches to an elevation of 465.6 feet.
Since we have a dewatering.well installed, to a.dd
'I E
conservatism to the design, we would be willing to. leave it r
in and operate.it when th'e water level got to elevation ju t as an added means of conservatism on the design.
).
- 476, s
I think that was our position and was in the letter we sent to the staff in August of this past year.
1 V.
[
i-
~
~
k 2169 082 E
b r
k I
s I
h.
I N.
F 4
[
E-F
g n.
f}
$5 h*
/
h
!=lisDDTDG? OU GBT 00N S'YDM )G GG G i P)/ d N.. i f
b :
../.4.,,W.2 i
A u
e h
- o.,,
By Dougfos Storr
. mildly radioactive water could con-first have been cleared with the de.
Earl llornm.ni, vare president o radioactiove malcrial.
rus rt.uticoortre crivaldy Icak, inspectors said, but sign firm.
3 o
" 8,
'Safrty proh! cms at the Zimmer probably not beyond the confines of.
sepports att he improperly anchored at CG&'t sa,mul electric Yin later found some of these engineering uclear st.itami m.iy delay the dant s the plant.
id the prohicms are weil y to being solved, lie sant. 'y.i i; tart up shfe, s(cording to ederal "This job is very Screwed tip."
to the concrcle walls.
on the t t.,
. :o tsiie. tors.
sai<1 isa T. Yin an inspcotur for the e In other cans. the desll;n firm many sufnds have hcen sW&
- N{!
I.uinly of Chicngn- '""h.,""'u[i1: y $ 'fi $ l Correcting the proldems will in. Nuclear Itegulatory Commission's -Sargent t-1
'O
'N
. hive reviewin!: rpecifications and Chicago ofhre. "This is a common never calculatc<l how much strain the supports wouhl take.
anchor expamlers-part of the cine. t i y. cih:ip= utennthemng some of.,the problem, but Zimmer seems to be h..u.sanda of wpgwo ts that ho d miles the worst case.
I o Yin further found that one of to. wall supports-and replaced 75,; ^
Itorgman said. The plant s designers.{ :!
q.; f 1.ipe throughout tho Moscow, Ohio OFFICIAI.S of the Cincinnati Gas the inspection programs at Zimmer have reviewed about half the pipe to-
.t r.
1.mt,
& Electric Co., the majority owner of -through which the problems should ceiling liangers. he said, and ] 'l s
," y!.L. Samenf ehose supgmits may not be the plant, smd the gyrobicms are have heen s potted-was inadcrpiate, prmuised to comp!cte the toh bCf.':
tr<ing enough, federal officials say, being corrected and said they do not o Finally, Yin noted that about rio April.
CG&E has @
M. nso..rtors pose that in addition to expect major delays.
of the plant s mare than twM asnish.
til; AnnEn that J
t in.hnr.vnn.? wpjwn t components too hispector Yin visited the plant hers"-anti shock desices that cush. orderrrt r,n new antiqtpx k smi!,lern,;t, '
- nk, th y foumi that, tlie calcula. four times since August 19731 et ion the pipes from carthr;unkes ', but wouhl wait to lest the old bnes ] -
' h ioneheck the streuth of the supports connected willt remitruction and.in-sound. Utility officials disputo this nmi inspections scepiired to 'which time he issued six citations were of a make he considered on.
before installing the n::w.
a.
Itorgm.m sani he exiwrts unfy a slight delay, if any, in the V.r,I and.4,j i.,,,; ecre not vampleteil.
~
'spection practices.
i opinion.
'UH FitMI'. til' T8tr.131.FS hivolved hon 6tntion's.fune 1971 inel lo.plinf.
"3
.T@ arry water erroli:d to. the safe.the CG&E six federal " infractions." duce c!cetricity sometime in 19).).
Ju his c.tations. Y.m noted thht TilESC Pn0Til, EMS carned date qThe station is ekpceted to pro, t i
i T hutilown of slu renetor, Itut federale g.rtors ::.of if the plant opened stra,ntiMy does not knmv how much
' C.
m the pipe supports will take.
Under law, on infraction is the sec. I!c said the coricetions will cost "a D :.
. f3 dtti these safety problems, and an e in trany cases he said, work-ond most serious of three possihlo lot-that s all I can say
! ? l. rcident ori urred. the reactor psib. crs did end install supports according viniatimts involving nuclear con.
But commission they expect a oclay closer to six.f g,
i tdy rpuld be safely shut down.
to the elenigners dranings. That i.1 struction.The most serious, a " viola.
'I 4
k[ round alter an acrident,couhl(fam. permitted but -any changes should. tion," involves the actual spillage of motibc.
ltrokrn pipes, however, whip ing '
i d
~ ^ - '
,.L.,
p
-e y, ne linjun iant capiipment at the r..
7-
,;t --
j d'
lant, federal s<nircen said, h ', * -
'. u iF 6
+
i r.;. In rare cases, pipes carrying [
4 I
n -
.I g {[
g '
. ;h
(
p e
3 p
q.m._.
4.;.
c..
z y
3
.....-..-.....v.--.-
- l s.a.
g Na Arrnewerrr co
'q t.AJ 4.
i