ML18047A316
| ML18047A316 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/03/1982 |
| From: | Lainas G Office of Nuclear Reactor Regulation |
| To: | Vandewalle D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV NUDOCS 8205120298 | |
| Download: ML18047A316 (33) | |
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Docket No. 50-255 Mr. David Vandet*Jalle OJ Nuclear Licensing Administrat Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201
Dear Mr. VandeWalle:
May 3, 1982 DIRIBUTION Docket NRC PDR L PDR ORB#5 Reading NSI C DCrutchfield HSmith
-n~ambach OELD OI&E ACRS (10)
SEPB
SUBJECT:
UNRESOLVED SAfETY ISSUES STATUS FOR PALI SADES PLANT The staff's safety evaluation report regarding the conversion of the Provisional Operating License for Palisades to a full-term operating license must address the status of unresolved safety issues (see discus~ion of ALAB-444 in Enclosure 1). To enable the staff to expeditiously review and evaluate the status of these items at Palisades, we will need up-to-date information of the type described in the enclosure to this 1 etter for the unresolved safety issues listed in Enclosure 1.
Accordingly, pursuant to §50.54(f)*o*f 10 CFR 50, you are requested to furnish the fol lowing information with regard to each of the identified unresolved safety issues within 60 days of the date of this 1 etter:
(l}
has the issue been resolved at Palisades; (2) if so, how has it been resolved; and (3) if full resolution has not occurred (including implementation of necessary hardware, procedures, etc.) what interim measures have been taken to assure that continued operation would not pose an undue risk to the public.
- This request for information relates only to the Palisades Plant and does
~ ~o/
not affect other licensees; therefore, OMB clearance is not required.
"" ~
82-05i2o298 s2oso3 PDR ADOCK 05000255 p
Enclosure:
As stated Sincerely,
/ I {.:'\\
-os-" ~s e r>u Original signed by Gus C. Lainas, Assistant Director for Safety Assessment Division of Licensing SEPB :C {JJfflr
- WRussel 1
- SEE PRsV U
.f-URRENCES 5/ 3 /82 cc w/enc-1 OFFICE~ ******s~~**11~x-e**
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NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO; 1961-335-960
Docket No. 50-255 Mr. David VandeWa11e Nuclear Licensing Administrator Consumers Power Company ~ _.
- ~ -~
1945 West Parnall Road Jackson, Michigan 49201
Dear Mr. Vandewalle:
.... ~~*
DISTRIBUTION Docket NRC PDR Local PDR ORB Reading NSIC DCrutchfield HSmith n1ambach OELD OI&E ACRS (10)
SEPB
SUBJECT:
UNRESOLVED SAFETY ISSUES STATUS FOR PALISADES PL~NT The staff's safety evaluation report regarding the conversion of the Provisional Operating License for Palisades to a full-term operating license must address the status of unresolved safety issues.. (See discussion of ALAB-444 in Enclosure 1). To enable the staff to expeditiously review and evaluate the status of these items at Palisades, we will need up-to-date infonnation of the type described in the enclosure to this le(:ter for the unresolved safety issues listed in Enclosure l.
Accordingly, pursuant to §50.54(f) of 10 CFR 50, you are requested to furnish the following information with regard to each of the identified unresolved safety issues within 60 days of the date of this letter:
(1) whether and how the issue has been resolved at Palisades; (2) interim measures that have been taken; and (3) why the issue has no safety implications pending resolution and should it not be resolved, alternative means to ensure that continued operation would not pose an undue risk to the puBlic.
This request for information relates only to the Palisades Plant and does not affect other licensees:; therefore~ OMB clearance ts not required.
Enclosure:
As stated OFFICE. """""CC**wtenC=J.
SURNAME ****** See.. nex+/-..
DATE.........................
NRC FORM 318 (10-80) NRCM 0240 Sincerely, Gus C. Lainas, Assista~~ Director for Safety Assessment Division of Licensing ure*:... V
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OFFICIAL RECORD COPY USGPO: 1981-335-96()
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Mr. David J. VandeW~lle cc M. I. Mi 11 er, *Es qui re Isham, Lincoln & Beale Suite 4200 One First National P)aza Chicago, Illinois 60670
- Mr. Paul A. Perry, Secretary
- Consumers Power Company
~12 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 CHERRY & FLYNN Suite 3700
.Three First National Plaza
.Chicago, Illinois 60602
- Ms. Mary P. Sinclair Great Lakes Energy A 11 i ance 5711 Summerset Drive Midland, Michigan 48640 William J. Sca~lon, Esquire 2034 Pauline Boulevard A~n Arbor, Michigan 481U3 Township Supervisor Covert Townshi Route l, Box 10 Yan Buren County, Michigan 49043 Office of the Governor (2) *
.. Room 1 - Capitol Building Lansing, Michigan 48913 Palisades Plant ATTN:
Mr. Robert Montross Plant Manager Covert, Michigan 49043
- e U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:
Regional Radiation Repr~sentative 230 South Dearborn Street Chicago, Illinois 60604 Charles Bechhoefer, Esq., -Chairman Atomic S~fety and Licensing Board
- Panel *
- U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. George C. Anderson.
Department of Oceanography University of Washington Seattle, Washington 98195 Dr. M. Stanley Livingston 1005 Calle Largo Santa Fe, New Mexico 87501 Resident Inspector c/o U. S. NRC Palisades Plant Route 2, P. o. Box 155 Covert, Michigan 49043 James G. Keppler, Regional Adminfstrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
ENCLOSURE l STATUS OF UNRESOLVED SAFETY IS_SUES AT.PALISADES The NRC staff evaluates the safety requirements used in its reviews ag~inst new information as it becomes ava~lable. Information related to the safety of nuclear power plants comes from a variety of sources including.experience from operating reactors; research results; NRC staff and Adviso~y Committee on Reactor Safeguards safety reviews; and vendor, architect/engineer, and utility design reviews.
Each time a new concern or safety issue-is identified from one or more of these sources, the need for immediate action *to ensure safe operation is assessed.
This assessment includes consideration of th~ generic implications of the issue.
In some cases, immediate action is taken to ensure safety.
In other cases, interim measures, such as modifications to operating procedures, may be sufficient to all ow further study of the is.sue before licensing decisions are*made.
In most case~,.*the initial aisessment* indicates that immediate licensing actions or changes in licensing criteria are not necessary.
In any event, further study may be deemed appr'opri ate to make judgments as to whether existing NRC staff requirements should be modified to address the issue for new plants or if back,fitting is appropriate for the long-term operation of.plants already under construction or in operation.
These.. issues are sometimes called 11generic safety issues" because they are*
related to a particular class or type of nuclear facility rather than to a specific plant*.
These issues have also been referred t9 as 11.un~e.solved safety issues (NUREG-0410, "NRC Program for the Resolu.tion of Generic Issues Related
~.
to Nuclear Power Plants," dated January 1, 1978).
- However, as discussed above, such issues are considered on a generic basis only\\ after the ~taff has made an
{nitia,- determination that the safety significance of the issue does not prohibit continued operat_io~_or require licensing actions while the longer-term generic review is under way.
A Decision by the Atomic Safety and Li cens i ~g App ea 1 Board -of the Nuclear Regulatory Cammi ss ion addresses these-*
longer-term generic studies.
The Decision was issued on November 23, 1977
I' (ALAB-444) in connection with the Appeal Board's consideration of the Gulf States Utility Company applicat1on for -the-Rtver Bend Station, Units 1 and 2.
In the view of the Appeal Board (pp. 25-29),
The responsibilities of a licensing board in the radiological health and safety sphere are not confined to the consideration and dispo~
sition of those issues which may have been presented to it by a party or an "Interested State" with the required degree of specificity.~
To the contrary, irrespective of what matters may or may not have been properly placed in controversy, prior to authorizing the issuance of a construction permit the board must make the finding, inter al°ia, that there is "reasonable assurance" that __
11the proposed facility can be constructed and operated at the proposed) ocat ion without undue risk to the health and safety of the public:"* Of necessity, this 10 CFR 50.35.(a) determination will entail an inquiry into whether the staff _review satisfactorily has come to grips with any unresolved
. generic safety problems which might have an impact upon operation of the nuclear facility under conside-ration.
The SER is, of course, the principal document before the licensing board which reflects the content and outcome of the staff's safety review.
The board should therefore be able to look to that document to ascertain the extent to which generic unresolved safety problems which have been previously identified in an FSAR item, a Task Action Plan, an ACRS r~port or elsewhere have been factored into the*staff 1s analysis for the particular reactor--and with what result.
To this end, in our view, each SER should contain a summary description of those generic problems under continuing study which have *both rele-vance to facilities of the type under review and potentially signifi-cant public safety implications.
This summary description should include information of the kind now.
contained in most Task Action Plans.
More specifically, there should be an indication of the investigative program which has been or will be undertaken with regard to the problem, the program 1s anticipated time span, whether (and if so, what) interim measures have been devised for dealing with the problem pending the completion of the investigation, and what alternative course of action* might be-avail-able should the program not produce the envisaged result.
In short, the board (and the public as well) should be in a position to ascertain-from the SER itself--witho~t the need to resort to extrinsic documents--the staffJs perception of the nature and extent*
of the relationship between each significant unresolved generic safety *question and the eventual operation of the reactor under_
scrutiny.
Once again,'th1s assessment might-well have a direct bear fog upon the _ab,iJ_ity of the 1 i censi ng board to make the s*afety findings required of it on the construction permit level even though the generic a~swer to the question remains in the offing.
Among*
- other things, the furnished information would likely shed light on such alternatively important considerations as whether:
2
(1) the problem has already been resolved for the reactor under study; (2) there is a reasonable basis-fQ'r.concluding that a satisfactory solution will be obtained before the reactor is put in operation; or (3) the problem would have no safety implications until after several years of reactor operation and, should it not be resolved by then, alternative means will be available to insure that.
continued operation (if permitted at all) would not pose an undue risk to the public.
This section is specifically included to respond to the decision of the Atomic Safety and Licensing Appeal Board as enunciated in ALAB-444, and as applied to an operating license proceeding Virginia Electric and Power Company (North Anna Nucle~r Power Station Unit Nos. 1 and 2), ALAB-491, 8 NRC 245 (1978).
In a related matter, as a result of Congressional action on the Nuclear Regulatory Commission budget for Fiscal Year 1978, the Energy Reorganization Act of 1974 was amended (PL 95-209) on December 13, 1977 to include, among other things, a new Section 210 as follows:
UNRESOLVED SAFETY ISSUES PLAN SEC. 210.
The Commission ~hall develop ~ plan providing for specifi-cation and analysis of unresolved safety issues relating to nuclear reactors and shall take such actions as may"be necessary to implement corrective measures with respect to such *issues.
Such plan shall be submitted to the Congress on 9r before January 1, 1978, and progress reports shall be includeq in the annual report of the Gommission thereafter.
The Joint Explanatory Statement of the House-Senate Conference Committee for the Fiscal Year 1978 Appropriations.Bill (Bill S.1131) provided -the following additional infor~ation regarding the Committee 1 s~deliberations on_~his portion of the bill:
SECTION 3 UNRESOLVED SAFETY ISSUES The House am1~ndment required development of a plan to resolve generic safety issues. - The* conferees agreed to *a requirement that the plan be submitted to the Congress on or before January 1, 1978.
The conferees also expressed the intent that this plan should identify and describe those safety issues, re 1 at i ng. to nuclear power reactors, which are unreiolved on the date of ~nactment. It should set forth:
3
f
~l) Commission ac~ions taken directly or indirectly to develop and implement corrective mea~ures; C?) f~ther actions planned concerning such mea:ur7s; and (3) timetables and cost estimates of such actions.
~he Commission should indicate the priority it has*assigned to each issue, and the basis on which priorities have been assigned.
In response to the reporting requirements of the new Section 210, the NRC*staff submitted to Congress on January 1, 1978, a report, NUREG-0410, entitled 11 NRC Pr_ogram for the Resolution of Generic Issues Related to Nuclear Powet Plants,."
describing the NRC generic issues program.
The NRC program was already in
~lace when PL 95-209 was enacted and is of considerably broader ~cope than the "Unresolv~d Safety Issues Plan 11 required by Sec~i.on 210.
In the letter trans-mitting NUREG~0410 to the Congress on December 30, 1977, NRC indicated that 11the progress reports, which are required by Section 210 to be included in future NRC annua 1 reports, may be more useful to Congress if _they focus on the specific Section 210 safety items.
11 It is the NRC 1s view that the intent of Section 210 was to ensure that plans were developed and implemented on issues with poteritially significant public safet~ implications.
In 1978, the NRC.undertook a.revi~w of more than.1$0.
- generic issues a~dressed in the*N~C prrigram to d~termin~ which i~sues fit this.
description and qualify as unresolved safety issues for reporting to the Congress.
The NRC review included the development of propo~als by the NRC
- staff and review and.final approval by the NRC Commissioners.
This review is described in NUREG-0510, "Identification of Unresolved Safety Issues Relating to Nuclear Power Plants - A Report to Congress," January 1979.
The report provides the following definition of an unresolved safety issue.
An Unresolved Safety Issue is a matter affecting a number of nuc1ear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants it affects.
Further,- the report :ind-icates that in applying this definition, matters that pose "important q,uestions concerning the adequacy of existing safety re qui re-ments11 were judged to be those for which resolution is necessary to (1) com-pensate for a possible major reduction in the degree of protection ~f the 4
r public health and safety or (2) provide a potentially significant decrease in the risk to the public health and safety.** "Quite s*imply, an unresolved safety issue is potentially significant from a public safety standpoint, a*nd its resolution is likely to result in NRC action on the affected plants.
All of the issues addressed in t~e NRC program were systematic~lly evaluated against this definition as described in NUREG-0510. *The issues are listed below.
Progress.on these issues was first discussed in the 1978 NRC Annual*
Report.
The number(s) of the generic task(s) (e.g., A-1) in the NRC program addressing each issue is indicated in parentheses following the tftle.
( 1 )
(2)
( 3)
(4)
(5)
(6)
(7)
(8)
(9)
(10)
( 11 )
(12)
( 1 3).
( 14)
(15)
(16)
( 1 7.)
( 18)
( 19)
UNRESOLVED SAFETY ISSUES (APPLICABLE TASK NOS,)
Waterhammer ~ (A-1)
Asymmetric Blowdown Loads on the Reactor Coolant System (A-2)
Pressur~zed Water Reactor Steam Generator Tube Integrity (A-3~ A-4, A-5)*
Anticipated T~ansients Without Scram (A-9)
Reactor Vessel Materials Toughness (A-11)
Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports A-12)"
Systems Interaction in Nuclear Power Plants ~A-17)
Environmental Qualification of Safety~Related Electrical Equipment (A-24)
Reactor Vessel P~essure Transi~nt Prot~ction (A~26)
Residual Heat Removal Requirements*(A-31)
Control of Heavy Loads Near Spent Fuel (A-36)
S~ismic Design Criteria (A-40) -
- Containment Emergency Sump Reliability (A-43)
Station Blackout (A-44)
Shutdown Decay Heat Removal Requirements (A-45.)
Seismic Qualification of Equipment in Operating Plants (A-46)
Safety Implications of Control Systems (A-47)
Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment (A.-48)
Pre_?s_uri zed Therma 1 Shock ( PTS) (A-49) the NRC staff has issued reports providing iti proposed resolution of six of these issues.
.5
NRC staff 1 s proposed re solution of s*i x. safety issues.
Task number A-2 A-9 A-24 A-26 A-31 A-36 NUREG report number and title NUREG-0690, "Asymmetric Slowdown Loads on PWR Primary Systems" NUREG-0460, Vol. 4, "Anticipat~d Transients Without Scram for Light Water Reactors 11 NUREG-0588, 11 Interim Staff Position on Envi-ronmenta l Qualification of Safety-Related Electrical Equipment" NUREG-0224, 11 Reactor VesseTcP-ressure Transient Protection for Pressurized Water Reactors 11 and NUREG-:oaoo, BTP RSB 5-2 SRP 5.47 and BTP 5-1, "Residual Heat Removal Systems" incorporate requirements of USI A-31 NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants With the exception of Tasks A-9, A-43, A-44, A-47, A~48 and A-49, Task Action Plans for the generic tasks above are included in NUREG-0649, 11Task Action Pl ans for Unresolved Safety Issues Related to Nuclear Power Plants." A technical resolution for Task A-9 has been proposed by the NRC staff in Volume 4 of NUREG-0460, issues for comment.
This served as a basis for the staff's proposal for rulemaking on this issue.
The Task Action Pl an for Task A-43 was issued in January l 981, *and the Task Action Plan for A-44 was issued in July 1980.
Draft NUREG-0577 which represents staff resolution of USI A-12 was issued for comment in November 1979.
Th~ Draft NUREG contained the Task Action Plan _for A-12.
T~e _in:f9_!matir;rn provided in NUREG-0694 meets most of the informational requirements of ALAB-444.
The Task Action Plan A-49 is attached.
Each Task Action Plan provides a deicription of the problem; the staff's approache~ to its resolution; a general discussion of the bases on which continued plant licensing or operation can proceed pending completion of the task; the technical organization involved in the task and estimates of the manpower required; a descr'iptfon of the interactions with other NRC offices, the Advisory Committee on Reactor Safeguards and outside organizations; estimates.of funding required for contractor-supplied technical assistance; prospective dates for completing the task; and a description of potential problems that covld alter the planned approach on schedule.
6
- .f In addition to the Task Action __ Plans, __ the ~s.laff issues the 11 0ffice of Nuclear Reactor Regulation Unresolved Safety Issues Summary, Aqua Book (NUREG-0606) on a quarterly basis, which provides current schedule information for each.of t.he unresolved safety*-_issues.
It _also includes information relative to t~e imple-mentation status of each unresolved safety issue for which techni ca 1 resolution is complete.
_7
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',f TASK ACTION PLAN PRESSURIZED THERMAL SHOCK (TASK A-49)
Lead Organization:
_ Task Manager:
Lead Supervisor:
NRR Principal Reviewers:
RES Principal Reviewers:
Applicability:
Projected Canpletion Date:
Division of Safety Technology *
(DST)
Roy Woods, Generic Issues Branch (GIB)
K. Kniel, Chief, GIB, DST B. Sheron, E. Thran, P. 0 'Reilly, A. Oxfurth, L. Lois, R. Johnson, R. Klecker, N. Randall,* W.
Haz~l ton, G. Vissing, J. Clifford, G. Schwenck C. Johnson, M. Vagins, P. Barana,.,isky, J. Strosnider, C. Serpan
- Pressurized Water Reactors May 1983
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- 1.
- INTRODUCTION AND BACKGROUND As a result of operating experience, it is now recognized that transients can occur in pressurized water reactors (PWRs) characterized by severe overcooling causing thennal shock to the vessel, concurrent with or followed by repressurization (that is, pressurized thennal shock, PTS).
In these PTS
. transients, rapid cooling of the reactor vessel internal surface causes a temperature distribution across the reactor vessel wall.
This temperature distribution results in thennal stress with a maximum tensile stress at the inside surface of the vessel.
The magnitude of the thennal stress depends on the temperature differences across the reactor vessel wall.
- Effects of this thennal stress are c001pounded by pressure stresses if the vessel.is repressurized.
Severe reactor systein overcooling events which could be followed by
.repressurization of the reactor vessel (PTS events) can result from a variety of taus es.
These include instrumentation and control System malfunctions, and postulated accidents such as small break loss-of-coolant accidents (LOCAs), main steamline breaks (MSLBs), feedwater pipe breaks, or stuck open valves in either the primary or secondary system.
As long.as the fracture resistance of the reactor vessel material remains relatively high, such events are not expected to cause failure.
After the fracture toughness of the vessel is reduced by neutron irradiation (and this occurs at a faster rate in vessels fabricated of materials which are relatively sensitive to neutron irradiation damage), severe PTS
- events could cause crack.propagation of fairly smal 1 flaws t~at are conservatively postulated to exist near the inner surface.
The assumed *initial.
flaw might initiate and propagate into a crack thrqugh the vessel wall of sufficient extent to threaten vessel integrity,~nd therefore core cooling capability.
The Rancho Seco event of March 20, l978 is believed to represent the most severe (and prolonged) ove'rcool ing transient experienced* to date *.
Although the event was* considerably less severe than would have been necessary to cause potential failure.of the Rancho Seco vessel at the time the event occurred (because of the existing fracture toughness of the vessel), the event nevertheless represents an important precursor for such severe events.
That is, had -subsequent failureS-or inappropriate operator actions or lack of proper operator actions occurred, the precursor that did occur could have 'developed into a more severe (but less probable) PTS event.
Similarly, had the Rancho Seco event occurred with a more highly irradiated vessel,\\ vessel integrity could have been jeopardized without the occurrence of additional.failures or errors.
In the Rancho Seco event, a lightbulb bein~ replaced in the non-nu cl ear ins.trumentati on/integrated control system {NNI/ICS) panel was dropped and caused a short to occur while the plant was at approximately 70% power.
About 2/3 of the instruments that indicate pressure, temperature
and level were lost.
Furthennore, the operator did not have confidence in the validity of indication or the remaining instrumentation.
The reactor tripped, feedwater was lost, the auxiliary feedwater (AFW) pumps started but remained isolated due to the JCS failure, and the once-through steam generators dried out.
Subsequent refilling by the AFW and possibly by the main feedwater (MFW) systems caused primary system overcooling and actuation of high pressure injection (HP!) and opening of the AFW isolation valves.
Actuation.of HPI and MFW caused severe overcooling rates (approximately 300°F /hr) until some of the pumps were shut off by plant operators.
Actuation of HPI also caused repressurization of the primary system.
Operators did not have what appeared to be a reliable temperature indication,
_and thus kept AFW and HPI on to maintain core cooling while restoring NNI.
During this time, primary system temperature had_ been reduced to about 28S°F ~
Since the March 1979 accident at Three Mile Islancf;"(TMI), much emphasis has been placed upon the need to run co9ling pumps until it is positively detennined.that they can be turned off without the possibility of core ovetheating.
Such training contributes to the severity of PTS events, however, and may be a factor in making future events of this type even more likely and/or more severe {the. Rancho Seco event occurred before TM!) *.
In view of the above, the program described in this Task Action Plan (TAP) is needed to fonnulate a regul~tory requirement to ensure that the risk of pressure vessel failure fran PTS events is sufficiently low through each vessel's design end-of-life. The program that will be conducted to provide finn bases for such a regulatory requirement includes:
development.
of methods for estimating the probability and severity of PTS transients and the operator's role in such events, refinement of methods for determining pressure vessel stresses in the event PTS transients do occur; _refinement of methods for determining material properties and failure vul nerabi 1 i ty of the vessel due to PTS stresses as a function of vessel exposure to neutron irradiation (and thus as a function of time in plant life);
- evaluation of potential benefits from potential corrective actions; and
. development of criteria for acceptability of plant safety margins under postulated PTS events.
This program will provide a benchmark to aid _NRC in assessing acceptability of several PTS studies currently underway in the industry, as well as fanning a basis for recommending acceptance criteria for resolution* of the PTS issue.
As stated in Section 3, {Basis for Continued Pla~t Operation *and Lf~ensing Pending Completion), up until the present time we have used a generic.
method for predicting vessel properties versus irradiation time and have concluded that. no event having a significant probability of occurrence
- could cause any pressure vessel to fail today or in the next few years.
However, using those generic methods {which are believed to be conservative) we _predict the necessity for SOOie type of corrective action before design end~of-life for: _several vessels.
The results of this program are needed to provide more detailed and realistic* {but still conservative) analyses-of systems r.es.p.ons~s, material properties, and risks before decisions are required regarding the nature and timing of the corrective actions.
A-49/2
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Potential corrective actions are discusse-d in Section 2.B.(7) below.
- They include ways to delay vessel embrittlement by reducing neutron fluence at the critical locations, ways to decrease the probability of PTS events with better control systems and/or operator actions, a way*to lessen the consequences of PTS events if they do occur (such as wanner injection water), and a way* to improve vessel properties (in-place -
annealing).
The magnitude of the problem described in this TAP with pressurized thermal shock was not appreciated during the design stage of currently operating
- PWRs, although pressure vessel thennal shock had been considered for many years in the context of assuring integrity of the vessel when subjected to cold emergency core cooling water during a large loss-of-coolant accident.
sa*sed on a series of thennal shock experiments (unpressurized) conducted at Oak Ridge National Laboratory (ORNL) beginning in 1976 which verified the associated fracture mechanics analyses, it was concluded that a postulated flaw would not propagate through the vessel wall during a large LOCA.
Therefore, the vessel 1s ability to contain water would be maintained during subsequent reflooding which would occur at relatively low pressure due to presence of the large break.
However, the possibility of concurrent or subsequent high pressure can negate the above conclusion and will be evaluated in the program described. in this TAP.
It should be pointed out that the NRC staff does not believe boiling water reactors (BWRs) have a significant PTS concern, for several reasons.
Most importantly, BWRs operate with a large portion of the water inventory inside.the pressure vessel at saturated conditions (that is, it exists as a mixture of steam and liquid water at the mixture's boiling temperature and pressure).
Any sudden cooling will condense steam and result in a
- pres sure decrease, so simultaneous creation of* high pressure and 1 ow temperature (necessary to cause a PTS concern) is very improbable.
BWR operating experience provides verification* that PTS events are very improbable.
Al thoug-h there have been numerous overcool ing events, there
- have been no significant PT*S events at any domestic or foreign BWR.
Also contributing to the lack of PTS concerns for BWRs is the lower fluence at.. the vessel inner wal 1, since BWRs have more water between the core and the vessel wall due to the recirculation flow path (water shields the vessel.frcm the core).
Finally, the operating pressure of BWRs is 1 ower, which al 1 ows the use of a thinner vesse_l wal 1.whi~h _re_s!JJ ts in a scmewhat lower stress intensity for a postulated crack.
- 2.
A.
PLAN FOR PROBLEM RESOLUTION General Approach to the Problem i;...
An outline *of the proposed integrated program to be conducted by the Office of Nuclear Reactor Regula~i9D (NRR) and the Office of Nuclear Regulatory Research (RES) utilizing the National Laboratories, with input from* industry including the PWR owners groups and_ eight_ selected utilities. is shown in Figure 1.
Throughout the program, NRC will obtain and utilize the advice of consultants who are canpetent in the various technical disciplines relevant to this program, including certain input from the Electric Power Research Institute (~PR!)
concerning thermal mixing.
Additionally, NRC will work closely with the Advisory Ccmmittee for Reactor Safeguards and its consultants.
A-49/3
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T A11 work perfonned through RES and at the National Laboratories will be utilized for input to the NRR licensing decision process, for use as appropriate (and if applicable). It is not the intent that NRR "censor" or overly restrict the course of the research programs.
Nor is it the intent that the conclusion of the research projects will be wholly incorporated into licensing requirements without modification.
NRR is responsible for dev-eloping licensing requirements, and will use the RES and National Laboratory results only as input to the licensing process.
T.he NRC program consists of the foll owing major sub-tasks. The first two tasks, designated as (a) and (b), are considered to be part of the short-tenn NRC program to be completed by about June 1982 and are not discussed.at length in this TAP which covers ttle.,:~,.ong-tenn program.*
Short-Tenn Program - Review of Industr.y Responses
- (a)
Review of infonnation requested by August 21, 1981 letter to industry groups and eight selected utilities. This will provide a reassessment of the PTS issue by about June 1982.
The reassessment will conclude whether or not there appears.to be a short-tenn (within approximately two years) significant problem at any operating plant and will recommend any corrective actions found to be necessary before ccrnpletion of the program outlined in this TAP.
Knowledge gained in these reviews will be utilized to guide the overal 1 NRC program (that is to emphasize work in the areas with the greatest uncertainty).
Details of this review can be found, for example, in TAC
- 47548 for H. B. Robinson, plus sequential TACS for the other seven plants involved.
(b)
Draft revision of the trend curves in Regulatory Guide 1.99,.
Revision 1'.
This revision will be drafted to reflect new*
surveillance data and the effects of nickel content on the predicted value of Charpy shift (that is, how irradiated material properties are detenni ned for certain pressure vessel materials).
Long-Tenn Program - Independent Analysis of PTS (1) Selection of PTS transients to be analyzed based -on *systems studies, human factors studies, and probabilistic and risk assessment analyses for three lead plants.
(2)
Selection, model improvement and verification of transient codes for use in calculation of the selected transients.
{3)
Calculation of.th_e_ pressure vs. time and the temperature vs.
tim*e of the water in contact with critical welds or base metal in the pres.sure.vessel for the selected PTS transients. (using the selected and verified codes).
A-49/4
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(4)
Improvement and experimental verification of a state-of-the-art fracture mechanics code to predict stresses and therefore crack initiation, propagation, and arrest for given pressure-temperatu.re*
histories at critical welds or base material, including consideration of wann prestress if aemonstrated to be applicable. This will include input from near-tenn fracture mechanics experiments perfonned by the Heavy Section Steel Techno~ogy {HSST) group at ORNL.
(5) Calculation of failure potential vs. irradiation embrittlement (that is, neutron fluence from the operating history) of the pressure vessels at the three lead plants for the selected PTS event sequences using the pressure and temperature vs. time histories from item (3) as input to the item (4) codes. These analyses assume pre-existence of a range of crack sizes infintely long. of*
various depths.
(6)
Perfonnance of sensitivity studies to determine changes in predicted vessel failure probability due to uncertainties in such parameters as copper content of the weld, initial crack size, lowest temperature of cooldown, etc.
(7)
Development of an understanding regarding feasibility of and benefits to be derived fr001 various proposed corrective actions, including revised fuel loading patterns to reduce fast neutron flux at the vessel wall, increased temperature of safety
.injection.water, improved control and instrumentation systems and/or operator actions to prevent repressurization, and vessel annealing.
(8) Development and publication of a NUREG report rec001mending a Regulatory *Positfon regarding PTS.inc.luding appropriate limits (if any) that-must be obs~rved at specific classes of plants, and potential corrective actions.
Each of these items constitutes a major sub-task.
Many of the sub-tasks are planned to proceed concurrently, but-sane must be sequential.
The accompanying Figure 1 is prpvided to show an overview. of the sub-tasks, including their relationship and schedule.
More details of each sub-task are given in the discussion below.
B.
Technical Content of Major Sub-Tasks (a)
Review of Requested lnfonnation Full details of this item, which:is part of the short-tenn review leading to a June 1982 reassessment of the PTS issue, can be founa in TAC #47548 for H. B. Robinson and sequential TACs for the,..other seven plants involved.
The item is summarized be1 ow.
A-49/5
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(b) e e
NRR has requested plant-specific infomation from eight selected 1 i censees regarding material propert-i es, operator procedures, and systems interactions that can cause PTS events and the probability of such events.
NRR will review this infomation (the "60-day" and 11150-day" responses to the August 21, 1981 letters to the eight licensees) along with other (generic) input from the three PWR owner 1s groups (and EPRI) to provide a reassessment of the PTS issue to the Canmission by about June 1982.
The reassessment-will -
conclude.whether or not there exists a PTS problem at any plant significant enough to warrant immediate corrective action, and will recommend those corrective actions, if any,.that must be initiated before canpletion of the program described in this TAP.
Knowledge gafoed fran these short-tenn reviews will be utilized as appropriate (for example, as a starting point) in the programs described in this TAP, and will
.guide the NRC program *to emphasize the. 0areas where the most unc*ertainty exists.
NRR has also initiated an effort through the Division of Human Factors Safety to improve operating procedures to lessen the probability of a severe PTS event.
The near-tenn program will result in ident1fication (by each licensee) of a recanmended method or "pathway" to avoid
- both overcooling events (with concurrent or subsequent pressurization) and overheating events.
Plant operating procedures will be put in place or revised as needed to facilitate the operator 1s task in maintaining plant safety, along with appropriate operator training in those procedures and their underlying technical basis.
Generic guidelines for updated procedures will be completed by mid-1982.
Plants that require immediate corrective action can have plant-specific procedures in place; and all training regarding those procedures complete by the end of 1982 if required to deal specifically with PTS events.
In addition, a task force has been fanned to audit procedures that deal with potential PTS events, and to audit operator training regarding those procedures* and regarding PTS phenomena.
These audits will be* canpleted for the eight selected pl ants by June 1982.
A second task force has
- been fanned to accelerate consideration of methods that could significantly reduce flux at the vessel wall.
A revised Regulatory Guide 1.99 will be drafted.
Based on preliminary analyses of the PWR surveillance data base, which was gathered as part of the thennal shock studies~ it appears that the fonnulas for the trend curves for Charpy shift in Regulatory Guide 1.99 should have a new nickel-dependent tenn included. This will be done in the draft Regulatory Guide.
The new tem will sharply reduce the observed overprediction when Regulatory Guide 1.99, Revision 1 is applied to low nickel material
. such as A302B steel. For high nickel material, the new tenn will have little effect.
In addition, the planned draft rev.ision to Regu,-a-tory Guide 1.99, Revision 1 will update-the data base and wilJ. pu_t the trend curves on a statistical basis from which both mean curves and upper bound curves will be derived.
A-49/6
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The remaining items discussed below are the long-term PTS program,*
the principal topic of-* this TAP.
(1)
Detennination of Event Sequences to be Considered Three major sub-tasks are involved in selecting the transients to be considered.
(1-a)Preliminary Development and Quantification of Event Trees for Transients Which Could Result in Overcooling.
NRR is perfonning a preliminary probability s~udy of PTS initiating events (precursors) including MSLB, large break loss-of-coolant accident (LBLOCA), small break loss-of-coolant accident (SBLOCA), core shutdown cooling by safety injection*
with flow out the pressurizer safety valves and no feedwater (such as, "feed and bleed" core cooling)~ and feedwater transients in which increased feedwater is supplied to the steam generators (SGs) canbined with steam flow out of the SGs through open dump or relief valves.
This study includes multiple failures and multiple operator errors. This study will be perfonned for three lead plants (one from each PWR Nuclear Steam Supply System vendor) selected as the
. optimum available combination* of typicality (vessel materials and control systems) and worst irradiation embrittlement. This stu'dy will incorporate information obtained in the responses to the August 21, 1981 NRC letters sent to eight representative plants.
(1-b)Development and-Quantification of Event Trees for PTS Events Including Review of Control and Safety Systems.
Results of item (1-a) will be input into a RES program with ORNL to Perfonn a study of detailed control -and safety system design at the three 1 ead pl ants. That contract is to provide details of control and safety system function$ and failure modes that may lead to PTS event sequences.
Owners of the three lead plants will provide to ORNL control, feedwater, and safety system functions pertinent to PTS event sequences.
ORNL will define about twelve event sequences in sufficient detail to provide input to Los Alamos
- National Laboratory (LANL) and Idaho Nuclear Engine~rir:ig -~g.boratory (INEL) calculations of reactor coolant pressure and temperature vs. time in the downcomer region. The event sequences specj.fied will include consideration of m~ltiple failures and multiple operator errors. Discussions will be held ~ith license~s of the three lead plants as PTS studies progress, and areas of disagreement
- between ORNL, the NRC staff, and th~ licensee (for example, credit for operator action or control system perfonnance and consideration of multiple failures) are to be indicated in the initial reports along with a justffication of the final position*.
A-49/7
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(1-c)Human Factors Studies An additional ORNL research project, managed by the Human _Factors Branch of RES, will address required operator actions for the
- transients being considered and result in an assessment of the probability and the effect of human errors on the likelihood of occurrence and severity of overcooling transients. The NRC will develop human error probabilities from this infonnation.
- The above results will be jointly used by NRC and ORNL to detennin~
which PTS events are the major risk contributors,.and these events will be used in sub-tasks 3., 5, 6 and 8 below (refer to Figure 1).
The results will also be used to review new procedures that will be adopted by PWRs to help prevent PTS ev~nts and to lessen the severity of those that do occur.
(2) Trans.ient Model Development a~d Verification
- Concurrent with sub-task 1, LANL and INEL will be developing and
- obtaining data to verify the TRAC and RELAPS, and SOLA codes which will be used to calculate P(t) and T(r,t) for the selected PTS events.
The three codes need sane model improvement and verification by comparison with data.
Code improvements are needed for the pressurizer model, for thennal mixing in the cold leg and downcaner regions, and to model the secondary (steam-feedwater) system.
Data on thennal mixing in the downcaner will be obtained frcxn an ongoing EPRI program and will be used to verify the SOLA code.
Brookhaven National Laboratory {BNL) will perform a QA function for the input decks and canpleted calculations.
(3) Calculation of P(t) and T(r,t)
These calculations will be performed at LANL and INEL for the Transient event sequences identified in sub-task 1 using the improved codes developed and verified in sub-task 2.
(4)
Improvements in Methods and Data for Fracture Mechanics Calculations Several different types of experiments are being planned or are underway to provide data needed for methods improvement.
These tests are planned as part of the HSST program at ORNL.
The experiments are d~sjg..oed to improve our understanding of flaw initiation, propagation, and arrest *
. so that fracture mechanics calculations will be more relevant to_.PTS conditions.
Planned tests include a series designed to further our under-.
standing of the wann-prestress phencxnenon and the limits of its app,.icability.
Ultimately it is hoped that the methods can be extended beyond the presently accepted linear elastic fracture mechanics methodology to include elastic-plastic fracture mechanics methods.
In particular, these.programs will focus on obtaining theoretical and emperical information on the effects of cladding-and the poterifi al benefits of wa nn prestressing.
Consideration wilLa_lso be give_n__t,ocr:ack propagation into material still on*the upper shelf, thus integrating A-49 with /\\-11.
A-49/8
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J Currently underway are a set of tests wtth*. small flaws in several
- square-foot, 2 inch thick plates that are stressed by four point bending (that is, no thennal or pressure stress). These tests will involve through-clad cracks, under-clad cracks, degraded cladding, and no cladding.
Later, irradiated samples will be used.
Also currently underway are a set of tests using cylinders approximately 3 feet in diameter and 4 feet long with various flaw geometrics which are tested using liquid nitrogen (but without pressure stress).
Some of these cylinders will be clad on the cooled surface to detennine cladding effects.
A oressurizerl thermal shock test is being planned which will be pressurized cylinder that will be thermally shocked to simulate*
both types of PTS stresses (thermal and pressure-induced).
Fracture mechanics codes (OCA-1 at ORNL and the NRC codes) will be
- further developed utilizing the above experimental results plus analytical work in the areas of: effect of cladding; treatment of through-clad cracks; treatment of warm prestress; three-dimensional effects; and size and shape of pre-existing cracks.
More precise fluence/materials data and properties information will be obtained and developed for use as input to these.calculations.
Results of this sub-task will remove known conservatisms where possible in the fracture mechanics codes.
(5)
Vessel Failure Analyses Cal culati ans will be performed using the methods and data from sub-task 4 and the P(t) and T(r,t) results fr001 sub-task 3 for PTS events.
This sub-task* s results wi 11 include the occurrence probabi 1 i ty of each PTS event from sub-task 1 and the consequences of each event (that is, crack initiation, propagation~ afrest, or through-wall penetration) at various times ifl the vessel life. These results will be used to provide a prediction of reactor vessel failure as a function of effective ful 1 power years (EFPY) of operation for the PTS events.
A r~nge of crack depths are assumed to pre-exist for these calculations.
Extension of any of those pre-existing cracks into a through-wall crack penetration will be assumed to produce vessel failure. Considering that sub-task 1 al so produced an-estimate of the frequency of each tran~ient considered, the last output of thi~-sub-ta~k ~ill bi-a 11best 11 estimate (somewhat conservative) of ve~sel failure probab.il i ty vs. effective full power years for the three (typical) lead plants.
These results will be condisered by NRR and useq as appropriate on one of the inputs into the licensing decision process.
(6)
Sensitivity Studies There are many uncertain.ties in the overall program -(sub-tasks 1 thr.ou_gh 5).
The_ ~ff~c~ of those uncertain_ties on sub-task _five's A-49/9
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results will be evaluated.
Examples are:
initial crack size, fluence and/or material properties, copper and nickel* content of the welds, temperature at the weld, cooling rate, and pressure.
Sensitivity of the program results described above to credible variations in these parameters (individual or varying in multiple canbinati ons simultaneously) must be assessed before a Regulatory Position can be detennined. This will be done in two diverse ways:
(a) A series of P(t), T(r,t) and fracture mechanics calculcations for several CClllbinations of different input parameters, will be perfonned to detennine the effects of variations in the input on outputs of sub-task 5.
(b)
NRG has developed a statistical, Mbnte* Carlo-based
. cClllputer code that will allow calculation of a response surface resulting fran a statistical variation of many input parameters. A statistical result can be obtained giving the mean value. of ri~ due to PTS events~ and variance in that risk, with consideration for the un*.
certainties.
Results of both methods will be utilized to arrive at a detennination of risk from PTS events at the representative three lead plants.
Since representative plants were selected, the results can in principle be generalized to obtain an approximate value for risk at other PWRs.
- Extrapolation, approximation, or engineering judgment may have to be used for specific pl ants that differ significantly fran the 11typi cal 11 1 ead pl ants selected.
(7) Benefits/Practicality of Corrective Actions Several potential corrective actions are possible, and will be considered. These include:
(a) Reducing the neutron flux at the pressure vessel. For example, sane of the outennost fuel elements in the core could be replaced with partially loaded or reflector elements or a fuel management program adopted that places partially depleted fuel elements. near the *vessel.* --
(b) Annealing the reactor pressure vessel in-situ to restore sane or all of the fracture toughness lost by neutron irradiation. Although annealing is feasible fran a metallurgical standpoint, and studies made to date have not revealed any damaging side effects, it would be e~pensive ~n~ _would require a long down time.
-. (c)
- Reduc_ing.. _th~ thennal shock during some transients by rais*ing the temperature of the emergency core cooling system injection water.
A-49/10
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(d.)
- Reducing the probability of *the event by new procedures, new control systems, new *instrumentation systems or a caTibination of all three to prevent repressurization*o.r give clearer indication to the operator th~t a situation is developing that has potential PTS concerns.
These corrective actions would provide automatic actions or allow operator actions with a higher degree of reliability to -
prevent repressurization.
The programs described below will provide the infonnation needed to assess the benefits to be derived fran, and the practicality of, the various proposed corrective actions.
ORNL will provide consultation to the NRC staff in evaluating the effectiveness of the various corrective actions as part of their ongoing contract with NRC.
In addition, BNL will evaluate effecti~eness of the fuel rearrangement or fuel removal corrective actions designed
. to reduce fast neutron flux of the vessel wal 1.
As part of licensee responses to the August 21, 1981 NRC request, the eight licensees have been asked to canment on the effectiveness and practicality of the various proposed corrective actions.
EPRI is s,ponsoring a program to evaluate the effectiveness of proposed corrective actions. They have already presented preliminary results of these studies regarding benefits to be derived from warmer safety injection water, and they have al so presented results of long-term benefits to be derived fran annealing irradiated pressure vessel materials at various temperatures, as well as a preliminary study by Westinghouse regarding the feasibility of in-place pressure vessel annealing. These res~lts were presented at the Ninth Water Reactor Safety Research Meeting, October 26-30, 1981, held at the N~tional Bureau of Standards in Gaithersburg, Maryland.
'f (8)
Regulatory Position Utilizing all of the above described information, particularly the risk vs. Ef PY fran sub-task 5 and the effectiveness of proposed fixes fran sub-task 7, the NRR staff will propose. a Regula:t;ory Position for Canmission approval and issuance for public and industry c001ment.
This proposed Regulatory Position wjll be ccxnpatible with the NRC 1s safety goal position currently under development.
After resolution of the canments, an implementation position will be recommended to the Canmission.
We anticipate that the implementation position will contain:
{lf required plant-specific limits; (2): suggested corrective actions for plants*that exceed those limits; and (3) a justification of the acceptability of pl ants.n.ot exceeding those limits.
A-49/11
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C.
Management of Work The responsibility for preparing and implementing a program to resolve this Unresolved Safety Issue is with the Generic Issues Branch, Division of Safety Technology (DST), Office of Nu cl ear Reactor Regulation.
A Task Manager in the GIB will provide overall management of all work identified in this Task.
Action Plan, including coordination of all work perfonned by other divi-sions-and branches; both within NRR and RES.
NRR will have the responsibility of taking licensing-related actions on pressurized thennal shock issues during the conduct of this program.
D.
Schedule The fo1 lowing schedule estimates have been develqped for the completion
- of the major ~asks of this program.
Tentative Schedule Sub-Task
{a) Review of Requested Infonnation (b) Draft of Revised Reg. Guide 1.99
{I) Detennination of Events (2) Transient Model Development
. (3)
P(t) and T(~,t) Calculation (4) Fracture Mechanics Code Development (5) Fracture Mechanics Calculation (6) Sensitivity Studies (7) Benefits of Corrective Actions (8)
Regulator.y Position Estimated Completion Date June 1982 June 1982 May 1982 May 1982 August 1982 September 1982 October 1982 January 1983 '
November 1982.
May 1983
. 3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLETION The staff has made a preliminary evaluation to detennine whether any immediate licensing action is necessary. This evaluation included: (1}
. the types of transients or accidents that could lead to overcooling of the* reactor system; (2) experience to date with. transient~ that have
.occurred at PWRs in the United States; {3) the probability that-such overcooling events will occur; (4) initial and irradiated material properties; and (5) the capability of reactor vessels to withstand these tran*sients.
based on fracture mechanics calculations.
Items 4 and 5 focused on the 1 i kel i hood of a fl aw exi stirig in a reactor vessel, material properties of the vessel, the copper content of reactor vessel welds, and the extent of reactor vessel irradiation (fluence).
A.
_Background Severe reactor system overcooling events which could occur under pressure -.
or be fol lowed by repressurization of the PWR reactor vessel (PTS.events) can result fran a variety of causes. These include instrumentation and A-49/12
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control system malfunctions and,_postulated accidents such as SBLOCAs, MSLBs, or feedwater pipe breaks.
Rapid cooling of the reactor vessel internal surface induces a temperature gradient across the reactor vessel wall.
The temperature gradient induces thennal stresses, with a maximum tensile stress at the inside surface of the vessel.
The magnitude of the thennal stress depends on the.
temperature differences across -the reactor vessel wall.
Effects of this
- thermal stress are canpounded by the pressure stress if the vessel is _
repressu rized.
As long as the fracture resistance of the reactor vessel material remains hjgh, such transients {except for extremely severe events) will not cause failure.
After the fracture toughness of the vessel is reduced by neutron irradiation, severe thennal transients could initiate crack
- propagation from fairly small flaws near the inner surface and result in significant cracki~g. The vessels of most concern are those with high radiation exposure in materials of relatively high sensitivity to raqiation.damage (such as those made with welds of high copper content) *
. For failure of the reactor pressure vessel to occur, a number of contributing factors must be present.
These factors are:
(1) a flaw of sufficient size to initiate and propagate; {2) a level of irradiation (fluence} and properties and ccrnposition sufficient to cause significant embrittlement of the material (the exact fluence is dependent upon trace elements present, that is, high copper content causes embri ttl emen*t to occur more rapidly); (3) a severe overcooling transient with repressurization; and (4) the crack must be driven to a size and location such that the vessel fails.
B.
Evaluation The staff preliminary review of overcooling events and their probabilities included a review of the staff's study on the. frequency of overcooling events at Babcock & Wilcox {B&W) plants {Ref. 1), a survey of operating experience on Westinghouse {W) and Ccrnbustion Engineering (CE) plants
{Ref. 2); a review of available accident analyses in Final. Safety Analysis Reports and in vendor topical reports; and a preliminary probabilistic analysis perfonned by DST (Ref. 3).
The preliminary results ~f these evaluations indicate that there is* a probability of about 10-per reactor year that a *s&W-designed plant will experience a severe overcooling transient similar to or worse than that experienced at Rancho Seco on March 20, 1978.
The Rancho s*eco transient was.the mos*t s*evere -overcooling trans!3nt experienced by ~ny PWR in the_Uni~ed.. :States.
This probability of 10 per reactor year includes contributions from steam generator control system malfunctions (the dcrninant contri~utor); SBLOCAs; main ~teamline or feedwater line breaks; and ccrnplete loss of feedwater flow.
The*staff estimated that the probability of such an overcool i ng event fn CE or W-designed reactors is lower, perhaps by an order of magnitude, than for B&W-designed *reactors.* This difference is based on design differences and-on operating experience.
~,._..
In the 1978 Rancho Seco transient, reactor pressure was ~aihtained at a -
fairly high level (1500 psi"g to 2100 psig) throughout the cooldown.
The A-49/13
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- minimum temperature of the reactor coolant (280°F) during the transient was. high enough so that material toughriess "of the reactor vessel was adequate. This evaluation leads the staff to believe that if this transient were to be repeated at Rancho Seco or any other B&W-designed facility within the next few years, the reactor vessel failure would still be unlikely. Nonetheless, the possibility of vessel failure as a result of an overcooling event cannot be canpletely ruled out.
If an overcooling event such as that at Rancho Seco were to occur, even for the vessel with the most limiting material properties in existance today, the staff would not expect a failure.
The staff conclusion is supported by the ORNL analyses of the Rancho Seco event (Ref. 4).
Reference 4 analyses and later ORNL analyses.
(Refs. 5 and 6) indicate that the threshold irradiation level for crack initiation (that is, small cracks gr011ing to l~rger ones assuming conservative initial material properties such 1
~s RTr-.mr=40°2 and copper
~ontent* of 0.35%) would be in the range of 10 neutron/cm
- The highest fluence to date in a B&W-designed facility is less than half the minimum value listed above.
It would, therefore,.be several years before any B&W-designed facility reached its threshold irradiation level.
Some reactor vessels in CE and W facilities have scmewhat higher fl uences; however, other mitigating factors--such as lower values of initial RTNRr--provide a significant margin of failure should an overcooling ev~ t similar to that at Rancho Seco occur.
I1;. should be pointed out that the NRC staff does not believe BWRs have a significant PTS concern, for several reasons. Most importantly, BWRs operate with a* large portion of the water inventory inside the pressure vessel at saturated conditions, (that is, it exists as a mixture of steam and 'liquid water at the mixture's boiling temperature and pressure).
Any sudden cooling will condense steam and result in a pressure decrease, so simultaneous creation of high pressure and low temperature (necessary to cause a PTS concern) is very improbable.
BWR operating experience provides verification that PTS events are very improbable since there have been no significant PTS events at any danestic or foreign BWR (that is, significant pressurization during or after a severe overcooling has not occurred). Also contributing to the lack of PTS concerns for BWRs is the lower fluence of the vessel inner wall, since BWRs have more water between the core and the vessel wall due to the recirculation flow path (water shields the vessel fran the core). Finally, *the operating pressure of BWRs is lower, which-results in a lower stress intensity at the bottan of a postulated crack.
C.
Conclusions and Recanmendations As a result of its evaluations to date; the staff has concluded that the probability of a severe-overcooling transient (similar in magnitude to the Rancho Sec.a*.~vent) is relatively 1 ow.
- Fo3 B&W-designed reactors this probability fs estimated to be about 10-per reactor per year, and for W-and,CE-designed reactors; it is lower, perhaps by an order of A-49/14
.. *.*,./... *.:... :.. -.-.. -* -**-... :... :.~*-*::.......... -.............--...
- ... ::........,.~,... _..,,._,_........., **------------
t T
magnitude.
In* addition, the st~ff has concluded that, based on present*
i rradi ati on levels at operating reactors, reactor vessel failure from such an event is unlikely.
Accordingly, the staff believes that no*
immediate licensing actions are required on operating reactors pending*
resolution of this issue.
For plants not yet licensed, licensing can proceed for all of the above reasons.
Also, the long-term PTS resolution will be produced by this TAP before irradiation history at those new plants is -
large enough to cause a significant PTS concern.
- 4.
TECHNICAL ORGANIZATIONS INVOLVED A."*
Generic Issues *Branch, Division of Safety Technology, Office of Nuclear Reactor Regulation Manpower Requirements:
1982 1-1/4 man-year 1983 1-1/4 man-year (See Section 2.C) - Overall coordination and direction of the effort will be provided by GIB.
B.
Office of Nuclear Reactor Regulation (Other Branches)
A signi.ficant portion of the work on this project will be performed. by contractors as discussed throughout this TAP and as summarized in Figure 1.
The contracts will be administered by RES, but the appropriate NRR personnel will be used to closely monitor and direct the various technical disciplines involved in the contract *work as* it progresses to assure that the ~ark produced satisfies the licensing needs.
In addition, several Technical Assistance programs will help with this work (see Section 4.D).
Also, the various contractor efforts (reports) will be reviewed when submitted.
Manpower estimates are given below in the fonn (x, y) where x is. the branch 1 s professi anal staff-year estimate for FY-1982 and y for FY-1983.
See also Table 1 for further summary of efforts involved.
The effort indicated on Figure 1 a~d in the paragraph below does not include the short-tenn PTS program described in items (a) and (b) above concerning the e.ight plants that received the August 21, 1981 letters, and the Regulatory Guid~ 1.99 Draft revision.
See TAC #47548 and the other sev~n sequential TAGS for the item (a) separate manpower request, or see the summary given in Table 1 of this TAP*which shows a line entry for each item.
The estimates and schedules below are for the 1 ong-tenn program described :in. this T~P. _
This TAP will involve:
the Materials.Engineering Branch (2, 2) (that is, 2 man-years in FY-1982, 2 man-years in FY-1983) for materials properties and fracture mechanics direction and support; the Probability and _Risk
- Assessment Branch (1/2,- 1/2) for support in the estimation of p.robabil ities for several PJS events and quantification :Of the event trees; the Reactor Systems Branch ( 1/2, V2); for direction of control system studies and transient code *development-and verification; the Instrumentation and Control Systems Branch (1/2, 1/2) for direction of control system studies and transfent code development and*verification; the Instrumentation and Control Systems Branch (l"/6, 1/6) for direction *of control system studies; Core.Perfonnance-Branch A-49/1 5
--~ -
~
- 1 (1/4, 1/3} for fluence studies and studies of-corrective actions involving fuel removal or re-arrangement to reduce flux at the vessel wall; the Division of Human Factors Safety (1/3, 1/3) for direction of studies on operator errors, procedures and training; and the Division of licensing {1/2, 1/2) for co.ordination of requests to licensees. A breakdown by branch showing when the manpower will be required is shown in Figure 2.
C.
Office of Nuclear Regulatory Research (2, 2)
RES resources will be utilized to administer the various contracts, and
- in addition they will provide consultations and guidance to the various technical review disciplines in NRR.
NRR is responsible for review.
milestones and licensing decisions, and time indicated for RES groups in this TAP are not to be construed as assignme.nts.
They are estimates*
of the time that will be spent as described*above *.
One of the two approaches to the sens.itivity studies will be perforined using methods developed by the Materials Engineering Branch of the Division of Engineering.
See description under sub-task 7 above.
The contracts will be:
ORNL will analyze event sequences leading to PTS and will estimate the probability of vessel failure at one "lead" plant for each PWR vendor.
LANL and INEL will im_p_rove and verify transient analysis codes and will calculate P(t) and T(r,t) for use in the ORNL fracture mechanics analyses.
BNL will study fluence to the pressure vessel and assist in evaluation of proposed corrective actions involving fast neutron flux reduction.
RES plans to participate in the EPRI/CREARE experiments tq obtain certain data needed for co~e development such as thermal mixing in the downccmer and cold legs.
Section B.b.4 describes the HSST program at ORNL that is also a part of the RES program being applied to the*PTS concern.
D.
Technical Assistance (also see Table 2}
The Reactor Systems Branch of the o;vision of Systems Integration, NRR will utilize Technical Assistance contracts at INEL and LANL to.r-eview several thermal hydraulics codes used by the licensees to calculate pressure and
- temperature history as a function of time for the selected event sequences.
The Core Performance Branch of the Division of Systems Integration, NRR will utilize *technical. assistance at BNL to benchmark the DOT 3.5 fluence cod~.
The Generic Issues Branch of the Division of Safety Technology, NRR will utilize a contract with Pacific Northwest Laboratory {PNL) to form a functional multi-di sci pl i nary group to investigate PTS.
The functi anal group wi 11 contain
- one or more experienced professional persons in:
probability and risk assessment systems (PRA), thennal hydraulics, materials, fracture mechanics-, and non-des-tructive examination.
The PNL effort will also utilize nationally known consultants in the various fields as necessary.
A-49/16
-.,.. *---. ~---------.::.~---------- --~-. - --~- -~ -~-- _,.. _________, __.. ____.... c,;.. -___ c __ ;, ___ *.,.. _:_ _ **~---~-,::~,... ;;_:_,. *.- ---. -_. _ _. __.. __,_. -..
__ --_,_....... -~~----~=--"'---~-. -- :_.,._._..:.,_.,_... :_.~-
e e
1
- 5.
POTENTIAL PROBLEMS A.
Close coordination and unity of purpose is required between NRR.and RES.
B.
Close cooperation is needed between ORNL and the licensees of the three "lead" plants.
C.
Close supervision of ORNL is needed fran a canbined 11NRR/RES" group.
D.
NRC and ORNL must see that LANL, BNL and INEL remain closely coordinated with the overall effort.
E.
Coordination and cooperation must be maintained. with industry to provide analyses and data for NRC studies.
A-49/17
(1
.I
'J REFERENCES
- 1.
"Insights on Overcooling Transients in Plants with the B&W NSSS,"
M. Taylor to S. F abi c, dated October 29, 1980.
- 2.
Nuclear Power Experience 1980, Bernard J. Verra, Publisher: Nuclear Power Experience, Inc. Encino, California.
- 3.
Frequency of Excessive Cooldown Events Challenging Vessel Integrity, A. Thadani to G. Lainas, dated April 21, 1981.
- 4.
Parametric Analysis of Rancho Seco Overcooli_ng Accidents, ORNL
- letter, R. D. Cheverton to M. Vagins (NRC, RES*), March 3, 1981.
- 5.
Ev al uati on* of Pressurized Thennal Shock, Oak Ridge Nati anal Laboratory, NUREG/CR-2083, October 1981.
- 6.
Staff review of ORNL Report on Pressurized Thennal Shock, Memorandum.
for the Canmissioners fran W. Dircks, EDO, October 30, 1981.
A-49/18
I.
.i A & B
- on ro ys em Safety System '--------------.
- Studies A & B
- uman actors!------------,
A & B tudies*
(1) Oetennination of
_.....;.._._... Events to Be Considered 5/82 (2) ransient Model 5/82
- Deve 1 opment; andi--~-------.....
---'M.~*:..:,_.:_;_:;~..:...=.;....;;.;;;........j Verification B.
- Plant Data.1-______
.. _~*-..
c ui sit ion (b ) ~raft Revis ion'-~ --~6i:.s..1.w.82.._ ______ _
~f R.G. l.*99 r A B & E (x) - Denotes major sub-tasks ~hich
- correspon~ to the text description.
.* - Involves item a) and denotes partial input from B--plant" responses.
Review
- of that material to be completed by 4/30/82* (Item a) of the Plan}.
Complete details of that review are in pl ant specific TAC--See for example TAC #47548 for H. B. Robinson.
A - NRC B - ORNL C - LASL
_A ty 1 83 egulatory 5 83
( 8} Pos 1 tion
- (7) enefits/
1118
~.
Practicality of1--__
- ..::..:..-...:..----------*--*---~*--------------'
"Fixes" FIGURE 1 LONG TERM PLAN i
- f.
~*
J
I.
r
.1
- 1 FISCAL YEAR 1982 FISCAL YEAR 1983 J
F M
A M
J J
P, s
0 N
0 J
F M
A M -~!
- ~ I
- n*
Completion,of
.. )
3 4
6 B :
Sub-Task #.:
I GIB: (1-1/4 PSY} 100% (relativ~ly uniform)
(1-~/4 PSY 100%
MTER ( 2) 30%
70%
(2)150%
50%
PRAB:
(1/2) 80%
20%
(1/ )
100% (relatively uniform)
RSB (112rsystems_., o es60%
~T Results 25%
15%
(l/f) 1xe 60%
( T -~) Events 40%
I CSB ( l /6 ve?~ ystems80%
( l /~)
~-
20%
100% {relatively uniform)
CPB (1/4) 20%
. 80% (fixles)
( l / ) 80%
20%
f
\\:'
I h
~.
I
{relatively uniform)
- 1.
DHFS (1/3)
Events 70%
30%
{l/f) 100%
I
- i.
Dk p /2}
100% (relativ@l uni form I
/!
100% (relativel uni form)
RES:
(2 PSY Total)*
I I
I I
Sensitivity_ Studies (1/2 PSY}*ll00% (relatively unifbrm) 80%
I 20%
Materials & Codes (FM) (3/4 PSY)*
60%
60%
T-H Codes (3/4 PSY)*
80%
20%
100% (relatively uniform)
Note:
Data shown are% of professional Staff Y~ars {PSY) time commitment - PSY shown in ( ).
~
=*See Section 4.C.
RES times indicated are estimates of consulting time and contract monitoring time that will be used but are not to be. considered commitments to the review effort aimed at generic licensing or Regulatory
. Retju i remeri ts.
Fl GURE 2.
.:.~~--...:~***:"....._:.;... :.: *........,;.. *_.*.,~~:........ : *. _:..: :..*.--.... _, ___ _..:..__:.. _..:..:..-:.....J.*.. *-* - -**-*--...... ~*--*-
- --*-***-**:.. _:~.--*:~---*-.*~......;.-.:..*---*-*---*.. ---;;..;:..... -... *:.:.* :~:.. :..
... :_..... -.~:..... '.-
- .* :... o *** :.-.************ c, -***-... ~* ---
'c l
\\
DESCRIPTION SHORT TERM PROGRAM (See Section 2.A.a and 2.8.a above)
TOTAL SHORT TERM LONG TERM PROGRAM (Reference Draft TAP for A-49)
TOTAL L9NG TERM RELATED PROGRAMS.
TABLE 1 PRESSURIZED THERMAL. SHOCK NRC PROFESSIONAL STAFF 'YEARS BRANCH OR FY 82 PERSON PSY DST/GIB 0.50 DST/RRAB 0.25 DL/ORB 0.25 DL/ORB 0.17 DSJ/RSB
. 0.25 DSI/CPB
- 0. 17 DHFS/PTRB
- 0. 17 DE/MTEB
. 0. 75 RES/MES 0.42 NRR 2.51 RES 0.42 DST/GIB 1.25 DST/RRAB 0.50 DL/ORB 0.50 DSI/CPB 0.25 DSI/RSB 0.50 DSI/ICSB 0.50 DHFS/PTRB 0.33 DE/MTEB
- 2.
RES/MEB 1
RES/Johnson 0.5 RES/Shotkin l
- 1.
. 4.58 RES 3.5 Reg Guide* l.99 Revisior.1-...
- _
. 0.25 P. Randall (See Section 2.A.b and 2.B.b above)
FY 83 PSY 0
0 0
0 0
0 0
0 0
.0 0
1.25 0.50
- 0. 50 0.33
- 0. 50 0.50 0.33
- 2.
l 0.5 l
5.60 2.67 0.2
.i TABLE 2
'PRESSURIZED THERMAL SHOCK - RESEARCH AND TECHNICAL ASSISTANCE I
FIN NRC FY K~tl)
F~$~~)
DESCRIPTION CONTACT CONTRACTOR B0119 HSST Vagins ORNL 4595 4677 B8133 U~R Pressure Boundary Integrity Va gins ENSA 500 600 85988 Surveillance Dosimetry Serpan HEDL 762 980 80415 Pressure Vessel S1mulation Ser pan ORNL 569 300 86224 Dosimetry Meas. Data Base Serpan NBS 128 200 87026 JR Cur~e Va gins USNA 60 70 A3215 Code Assessment and Application Shotkin BNL
~
A6047 Code Asses$ment and Applications Shotkin INEL 800 700" A7027 Analytical Res. ih LWR Safety Shotki n U\\NL A7217 TRAC Cale Assistance Shotkin LJl.NL 80468 Pressurized Thermal Shock C. Johnson ORNL 500 300 A7272 Reactor Systems Support of Operating Reactors Action Item Throm LANL 235 100 I B0763 Review of LOFTRAN and MARVEL Throm ORNL 35
- o.
A338l Pressure Vessel Irradiation Embrittlement Lois BNL 180 200 USI A-49 at PNL (Review group and individual consultants)
R, !foods PNL
- 400 400 TOTAL - PTS PROGRAM
. 8764 8527 I
l (1)
Doll~rs shown are for the por.tion of*tfle flN which is for PTS. The only two FINS which are exclusively PTS are B0468 and the Undesignated FIN for USI A-49.
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