ML20030A928
| ML20030A928 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/27/1981 |
| From: | Vincent R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-06, TASK-3-6, TASK-RR NUDOCS 8107300012 | |
| Download: ML20030A928 (11) | |
Text
e w.
-e Consumers Power Company Generei oMices: 212 West Michigan Avenue Jackson, Michigan 40:P01 + (517) 7884550 July 27, 1981 o
is.
Director, Nuclear Reactor Regulation D
h' f / h(
Att Mr Dennis M Crutchfield, Chief Operating Reactors Branch No 5 d
g o g 19SN.g U S Nuclear Regulatory Cor::=ission Washington, D C 20555 DOCKET 50-155 - LICENSE DPR g.
pf p
BIG ROCK POINT PLANT - SEP TOPIC III V-
- ( g g SEISMIC DESIGN CONSIDERATIONS A raismic reanalysis program for the Big Rock Point Plant has been the subject of nu=ercas letters and discussions since January,1979 Throughout 1979 and 1980, a creat deal of work was done, and a najor reanalysis effort was under-taken to seismically analyze important plant structures, and the primary cool-ant system.
The cost of this work, to date, has been well over one million dollars. Con-current with this reanalysis effort, Consumers Power Company developed a full Probablistic Risk Assessment (FRA) of the Big Rock Point Plant which included consideration of the seismic hazard at the plant site. The conclusion reached by the FRA was that the risk from a seismic event was not a significant contri-bution to overall plant risk. In light of the PRA conclusion that seismic risk was not significant, and the major cost of reanalyses to current criteria, Consumers Power Cc=pany proposed by letter, dated June 19, 1981, tnat the PRA be used with appropriate revisions to answer the balance of SEP Topic III-6.
The overriding considerations behind this proposal vere: 1) that Consumers Power Cc=pany had concluded that the Big Rock Point Plant, as it exists, is a safe plant which does not represent a significant risk to the health and safety of the public; 2) that the seis=ic contribution to the already 2 7v overall plant risk was insignificant; and 3) that continued expenditures strictly in support of the traditional NRC deterministic licensing process for those items consider-ed to be of little significance could not be sustained without making continued plant operation uneconomic. This approach met with a great deal of opposition from the staff. Criticisms of this approach centered around the lack of specific deterministic analyses to support assumpti,ans about plant response to a seismic event; questions concerning the FRA treatment of equipment fragility data, etc which had been developed from other studies; and a general skepticism that a probablistic approach can ever be acceptable for the treatment of external events such as earthquakes. As a result, we have conclude,1 that additional
~ 8107300012 810727 (0fg DR ADOCK 05000
I Dennis M Crutchfield, Chief 2
U S Nuclear Regulatory Commission Big Rock Point Plant July 2T, 1981 analyses vill be performed to deterministically prove that needed structures and systems will not fail in the event of an earthquake near the Big Rock Point site. The purpose of this letter, therefore, is tc discuss our proposed seismic a
reanalysis program for the balance of SEP Topic III-6,to provide a brief descrip-tion of work completed to date, and to discuss in more depth the reasons why continued operation of Big Rock Point prior to completion of the deterministic system analyses does not repres,ent a significant hazard to the health and safety of the public.
As discussed abov's, Consumers Power Company initiated a reanalysis program in 1979 to prove the seismic resistance of major plant structures, the primey coolant system and portions of some other systems. The scope of this program included the following systems / subsystems:
r cor building internal structure
.tainment shell Primary coolant loop Turbine building Service building and office addition Reinforced concrete stack Sphere ventilating room N al cask leading dock / core spray equiptent room Screenhouse/ diesel generator reen/ discharge structure i
Intake structure Suried Fire Main Piping Lake bed pipe Underground electrical cable 3uried fuel tanks Liquid radvaste vault l
These analyses were performed to a Regulatory Guide 1.60 spectrum he.ving a zero period horizontal ground acceleration of 0.12g.
This earthquake is some-(
vhat mora conservative chan that developed by the NRC for the Big Rock Point site as transmitted to Consumers Power Company by letter dated June 8, 1981.
The results of these analyses shoved that all items were verified to be adequate j
for the postulated ear +hquake (using service Level C stress limits) with the exception of the service building (s=all number of minor bracing modifications were reco== ended as being more cost effective than additional analyses) and prf wy coolant syste= junctions between the k" crosstie and the 2h" recircu-lation lines (conservatisms in stress intensification factors used and simplifi-ed modeling technique are believed to be responsible). On June 9, 1981, consu=ers Power Company infor: ally provided the NRC with a copy of the draft final report for these analyses. Since that time, our internal reviews of the report have been completed, and discussions are underway with our consultant to resolve ec==ents.
i'e expect to have this report ' finalized and formally submitted to the NRC in the near future.
1
Dennis M Crutchfield, Chief 3
U S Nuclear Regulatory Commission Big Rock Point Plant July 27, 1981 For the balance of SEP Topic III-6, Attachment,A provides a description of our planned seismic reevaluation program for Big Rock Point. This program is consistent with the scope defined in NRC letter dated April 24, 1981, and has been discussed with the NRC at a meeting on July 15, 1981.
It will be noted
$ hat system evaluations vill, wherever possible, rely on simplified analysis techniques and qualification by comparison with equipment which has previous-ly been qualified or has survived past earthquakes. During the July 15, meeting, some questions were raised about the level of detail being provided for the eval-uation criteria, and it was requested that the criteria be further amplified.
Subsequently, in a telephone discussion between R Hermann (NRC) and R A Vincent (CPCo) on July 22, 1981, Mr Herman agreed with the approach that the staff first review our program in more detail and provide specific questions in areas for which more infornation is needed. It will be our intent, therefore, to provide additional information in response to specific staff questions rather than to unilaterally supplement the information being provided herein.
Consumers Power Co=pany re=ains convinced that the hazard to the health and safety of the public, as a result of a seismic event, is not significant. This conclusion is based on the following:
i 1.
The Probabilistic Risk Assessment for the Big Rock Point Plant submitted to the NRC by letter dated March 31, 1981, shows the overall risk from the plant to be lov.
Plant risk is predominantly controlled by the plant size (available source term), probability of fission product release, energy available for release and population distribution. Since Big Rock Point is a s=dl plant, remote from =ajor population centers, even =ajor core-meltaccidents with containment failure vculd not result in any predicted acute fatalities, and only s=all numbers of latent cancer fatali-ties (small enough to be of questionable statistical significance) over the forty years subsequent to an accident. This is directly supported by the siting study recently performed by Sandia Laboratories for the NRC.
j Consumers Power Company has a stronger interest in preventing a =ajor accident than even the NRC, and can certainly not endorse conditions which would allow a =ajor accident to happen, but Big Rock Point Plant can and should be regulated in a mnner which recognizes the difference in risk between a Big Rock Point and larger, more complex plants.
2.
As discussed in our letter of October 10, 1980, the Big Rock Point Plant is in an area of very lov seismicity. The design earthquake with a return period of 1,000 to 10,000 years has been determined to be on the order of
.08g in the vicinity of the Big Rock Point site. The acceleration deter-mined by the NRC, as being acceptable in your letter of June 8,1981, is anchored at 0.llg.
Typical industrial construction is not usually damaged by earthquakes of these levels. Experience and other analyses performed to date have shown the Big Rock Point Plant to be generally a well-designed, well-constructed facility.
a 3.
The general conclusions from both Consumers Power Company and NRC seismic experts, who have inspected the Big Rock Point facility, also support the
Dennis M Crutchfield, Chief h
U S Nuclear Regulatory Commission Big Rock Point Plant July 27, 1981 inherent seismic resistance of the plant. Although some portions of plant systems were judged to be =arginal, none of the experts predicted failures which would cause loss of system functions. This is particularly significant in that this group of experts included at least one who has extensive experience in inspecting da= age to commercial facilities result-ing from earthquakes =uch = ore intense than the postulated earthquake for the Big Rock Point site.
h.
With respect to the question whether a seismic event could induce a LOCA, concern has been expressed about s=all lines attached to the primary cool-ant system which have not been inspected or analysed. This concern was raised because seismically induced deflections in the worst location have been predicted to be up to 3.4 inches. To put this value in perspective, it =ust be noted that PCS thermal deflections from heatup and cooldown can be up to two to three inches in the vorst locations. These deflections have been routinely experienced =any times without failure over the past twenty years of plant operation. These thermal deflections were identified in the original analyses and vere accounted for in the piping design. This understanding, along with a review of IE Bulletin 79-14 voc, supports the position that the small lines attached to the PCS piping at points of larp deflection are not stiffly supported and thus can sustain large deflections.
In addition, the cc=puter model used to predict this maximum dr.riections is a simplified =odel v'11ch does not account for the co=plete restraint configuration on the piping. As a result, we believe that this predicted value for =aximum seismic deflection of PCS piping for the assumed.12g earthquake is conservative. This conclusion is further underscored in that the spectrum used in the analysis is considerably higher at the relevant frequencies than the site-specific spectrum developed for Big Rock Point Plant by the NRC.
5 With respect to tne plant's capability to make up water to the PCS and the emergency condenser, the fire =ain is important in that it supplies primary and backup core spray, and provides a backup source of water for the shell side of the emergency condenser. Although none of the experts predicted failure as a result of the postulated earthquake, we recognise the staff's concerns with the threaded fire syste= piping. Redundaney does exist, however, to the fire system.
For makeup to the emergency condenser, the primary source is deminerali2.ed water. This is supplied through quality lines of all velded construction from the DW storage tank. If a seismic event caused a loss of offsite power, the DW pump and an air ce= pressor could be powered from the emergency diesel for makeup.
If the EW tank becomes depleted, it could be supplied from the domestic vater system onsite with the water sources being the do=es-tic water accumulator, well water storage tank via the do=estic water pu=p, or ultimately the deep vell pump. More than sufficient pumping capacity exists to makeup at the 10gpm (approxi= ate) rate needed for decay heat removal.
_~
Dennis M Crutchfield, Chief 5
s U S Nuclear Regulatory Commission Big Rock Point Plant July 27, 1981 As previously stated, an additional source of makeup to the emergency condenser is the fire eater system. The fire water makeup line taps off the core spray piping inside contai.utent through a manually-operated valve.
If na threaded fire water piping in the turbine building should fail due to a su event, further redundancy is provided by the Post Incident System. The Post Incident System can supply water to the core spray lines via a separate line from the yard loop through motor operated valve MO-7072. This system is also a high quality system of all velded construction. MO-7072 can be operated by hand or with power from the emergency bus. Since previous analyses have verified the seismic adequacy of the yard fire piping, and several valves are provided around the loop to isolate leaking sections, this would be a reliable water source.
With respect to makeup water for the PCS, none would be required for a long period of time if a LOCA did not also occur as long as a closed heat transfer -
loop such as the emergency condenser is available.
If a 10CA did occur, a large volu=e of makeup water vould be required. This water would normally be protided through the core spray or redundant core spray lines from the fire system. Again, failure of _ the threaded piping in the turbine building would -
not preclude core spray flow because of the redandant path which exists through the Post Incident System. Further redundancy is provided by the condensate pump. An estimated 20,000 gallons of water is norrally available in the hot well and the condensate storage tank as well as spproximately an additional 3,000 gallons in the demineralized water tank. This source vill assure core reflood for a top break (ie, above the core). Domestic water can also be used to supply water to the hot well for long term decay heat removal.
As noted above, the seismic adequacy of the screen house and the intake structure have also been verified. While the above scenarios encompass a loss of screen house equipment, failure of lines in the screen house would not be expected.
Since most items including the fire pumps are essentially at ground level, l
bui2 ding amplification of the lov 2eismic input accelerations would be small if it occurs at all. For the fire piping, in particular, since the runs in the screen house are relatively short and flexible, fa# N e is also not probable.
i l
In s m ry then, Consumers Power Company believes that the lov accelerations associated with earthquakes in the vicinity of Big Rock Point combined with the 4
infrequency of such events is sufficient in itself to allow continued operation while the remaining seismic analyses are completed. Analyses to date continue to verify the seismic resistance of systems and structures installed during original plant construction. As further assurance, however, the above discuss-ions have shown that redundancy does exist so that the plant is not solely reliant i
l on the fire water system (judged to probably be the most fragile from a seismic i
f l
l L
l
4 Dennis M Crutchfield, Chief 6
U S Nuclear Regulatory Commission Big Rock Point Plant July 27, '1981 standpoint) for makeup to tb emergency condenser and the pri=ary coolant system. We believe, therefois, that further actions are not warranted over che period defined in Attachment A for the balance of the seismic e.nalyses.
. lC t U
Robert A Vincent Staff Licensing Engineer CC Director, Region III, US iRC liRC Resident Inspector - Big Rock Point 4
1 l
i 1
I s
4 E
J t
1 k
?
BIG ROCK POINT PLANT Seismic Re-evaluation Program i
i 1
t Attachment A
I.
Purpose:
1.
To satisfy needs of Systematic Evaluation Program topic III-6 Seismic Design Considerations 2.
To provide a deterministic basis to better support assumptions and conclusions regarding seismic contribution to Plant risk ss developed in the Big Rock Point Probabilistic Risk Assessment II. Scope: Evaluations will be performed for the following systems or portions thereof:
a) Reactor Coolant System b) hin Steam up to and including MSIVs c) l'aedwater system up to and including isolation valves d) Portions of other systems directly connected to the RCS up to and including isolation valves e) Control Rod Drive system l
f) Emergency Condenser including makeup water piping g) Fire system including core spray, backup core spraf, enclosure spray, and backup enclosure spray h) Reactor Depressurization system i) Post Incident Cooling system j) Emergency Power Supply k) DC Power system
- 1) Spent Fuel Pool m) Appropriate portions of structures housing the above system Note that this list is essentially identical to the list provided in NRC letter dated April 24, 1981. This list is also essentially identical (with the exception of the instn: ment air system *) to the list of critical systems determined by the PRA to be necessary to mitigate all postulated core melt sequences.
- Instrument air is presently needed only for demineralized water makeup to the shell side of the emergency condensor. The FRA has identified replacement of a one inch manual valve on the redundant fire water makeup line to the emergency condensor as a desireable modification. This modification wculd allow the operator to supply fire water to the condensor in the event of failure of the nomal demineralized water makeup source. Instrument air, therefore, is not included on the above list.
III. Program Elements 1.
Compile existing BRP plant seismic qualification data for:
- Structures
- Critical Systems & Components
- Block walls & noncritical equipment presenting potential hazards due to proximity to critical systems.
2 2.
Complete review of D'Appolonia report
- Modify report to reduce conservatism if appropriate.
- Use report results and procedures as input to preliminary evaluation criteria.
- Docket revised report or existing report with qualify-ing comments.
3 Develop preliminary evaluation criteria.
- SEP infomation & applicable NUREGs, Reg.
Guides, etc.
- BRP FHSR and equipment specs.
- Service Level D vs C
- Applicability of & procedures for system walkdown.
4.
Conduct detailed, systematic plant inspection of critical systems and components.
- Identify vital c:m:ponents on complex systems such as control panels.
- Identify other items potentially hazardous to safety related equipment in their immediate vicinity (e.g.
block walls).
- 0btain necessary dimensions, etc. to supplement existing documentation.
- Photograph all items to be reviewed.
- Judge seismic fragilities of all components using pe:sonnel expert in observation of actual earthquake damage.
- Cbserve adequacy of bracing and anchorages.
- Identify most fragile lines & components. Detemine and implement desirable plant changes to judgment criteria for housekeeping items plus small lines below sizes considered by applicable codes.
- Take sample measurements of equipment responses to determine fundamental frequencies (hand excitation, accelerometer, oscilloscope).
- Separate items inact essible during plant operation for walkdown during outage.
- Target completion date is 9/30/81.
5 Finalize evaluation program. Eelect the best option for evaluation of e ch component based on available data. Options include qual ~ication by similarity, test, simplified analysis, detailed ana', s, replacement, sampling, etc.
{
6.
Prioritize BRP systems for evaluation acceding to:
- Criticality of system to plant operation or shutdown including degree of redundancy General ranking vill be RC pressure boundary, shutdown systems, ECCS rehted systems in that order.
3
- Perceived fragilities based on walkdowns.
- Accessibility during operation.
7 Finalize evaluation criteria to be applied to each system and method, to include ground and floor response spectra, damping, simplified analysis methods, etc.
8.
Conduct evaluation of each included component or piping system by one or a combination of the following methods:
a) Simplified Analysis (systems, eqttipment)
- Conduct single-degree-of-freedom analysis where justified, (e.g., compact structures whose fundamental mode is obviously rocking),
compute fundamental frequency to determine the evaluation g-level from floor response spectra, check the anchorage capacity.
- Conduct simple finite element analysis where justified.
- Fomulate conceptual designs of added bracing and anchorages where required.
Detailed Analysis (systems, here other) analysis equipment b)
- Conduct dynamic anklysis w methods can not be justified.
- For=ulate conceptual designs of added bracing
& anchorages where required, c) On-Site Testing (equipment)
- Perfom low-level excitation tests on BRP equipment to measure model frequencies, mode shapes, model dampings.
- Use test results to estimate participation factors and hence maximum response frem floor spectra without need of equipment structural detail for modeling.
- Use test results of BRP equipment to compare with similar equipment at data source plants, show that similar more fragile equipment have withstood seismic events.
- Use cessurements of damping as input to criterir,to justify more realistic assump-tions for evaluation.
d) Qualification by Similarity (equipment)
- Compare BRP equipment to similar equipment in data source plants which withstood seismic events. Show by analysis or testing that data source equipment is either similar or more fragile than BRP equipment.
- Detemine as necessary the fragility levels of BRP equipment (maximum tolerable g-level, or preferably qualification spectrum) from existing shake table data for similar equipment.
e) Target date for completion is the end of the 1982 Big Rock Point refuel.ing outage (1st quarter of 1982).
1 9
Develop Summary Report 10.
Consider interim actions (e.g. temporary systems for added redundancy) where necessary for items of particular safety significance.
11.
Complete any modifications determined to be necessary on a schedule which will allow completion with plant or other CPCo labor. Target,
date is end of 1983 Big Rock Point refueling outage although the final date will depend on the scope of the modifications identified.
.-.- - --....,.... -. -