ML19275K610
| ML19275K610 | |
| Person / Time | |
|---|---|
| Issue date: | 02/09/1982 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Bradford P NRC COMMISSION (OCM) |
| Shared Package | |
| ML19275K611 | List: |
| References | |
| FOIA-82-167, RTR-NUREG-0651, RTR-NUREG-651 NUDOCS 8203090708 | |
| Download: ML19275K610 (15) | |
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DISTRIBUTION NRC Central SHanauer WDi rcks, Cornell WMinners Rehm/EDO MErnst EDO #11423 SubJ File FEB 0 91982 d9 Schilk RDeYoung, IE FRemick Cavanaugh RMinogue LBickwit PFine CMichelson MEMORANDUM FOR:
Commissioner Bradford BSnyder RVollmer RMattson HThompson FROM:
William J. Dircks PCheck DEisenhut Executive Director for Operations Denton/ Case
SUBJECT:
STATUS OF RECOMMENDATIONS MADE IN NUREG 0651, " EVALUATION OF STEAM GENERATOR TUBE RUPTURE EVENTS" This is in reply to your Inquiry dated January 22, 1982, concerning the status of the recommendations made in NUREG 0651.
Since the technical issues raised in the report and its recommendations were not considered to be of immediate safety significance, the staff has been and is considering these recommendations in the context of other related safety programs.
The 16 recommendations for industry action (Enclosure 1, items 1.A-D) are all being addressed as part of related ongoing programs.
Details are provided in the Enclosure.
The six reconunendations directed at the staff (Enclosure 1, items 1.E-J) are subsumed into the program for development of improved Emergency Operating Procedures (ltem I. A) and post-TMI improvements in training, or are parallel to the recommendaticns for industry action.
As a result of the incident at Ginna, we are reviewing our requirements relevant to steam generator tubes and reconsidering the priori ve afforded this issue.
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& Rehm/EDO M. Ernst EDO #11423 Subj. File EDO Rdr PPAS S. Chilk R. DeYoung, I&E F. Remick R. Minogue L. Bickwit C. Michelson R. Vollmer H. Thompson D. Eisenhut Denton/ Case MEMORANDUM FOR: Comissioner Bradford R. Mattson P. Check FROM:
William J. Dircks B. Snyder Executive Director for Operations P. Fine S. Cavanaugh
SUBJECT:
STEN 1 GENERATOR TUBE RUPTURE This is in reply to your inquiry dated January 22, 1982.,
While we have not replied explicitly to the recoamendations in NUREG-0651, they are being addressed in other programs. The enclosed Staff Report gives the details.
The 16 recommendations for industry action-(Enclosure 1. Items I.A-D) are all being addressed as part of ongoing programs.
The six recommendations directed at the staff (Enclosure 1. Items I.E-J) are subsumed into the program for development of improved Emergency Operating Procedures (Item I.A) and post-TMI improvements in training, or are parallel to the recornendations for industry action.
As a result of the incident at Ginna, we are reviewing our requirements relevant to steam generator tubes and may revise them.
William J. Dircks Executive Director for Operations
Enclosure:
Staff Report cc: Chairman Palladino Cornissioner Gilinsky Commissioner Ahearne Commissioner Roberts SECY D/ DOE D/DL PE RVollmer DEisenhut GC 2/ /82 2/ /82 CONTACT:
S. Hanauer, DST /HRR
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Raff Report by Division of }M.y Technolo 7y,, NRR Status of Recommendations pf ating t6 steam Generator Tube Rupture Events l
February 2,1982 s'
I.
Recommendation in "Ev31uation of Ste'am Generator Tube Rhture Events,"
A.
Emergency Operating Procedures (Recommendations fotijicensees #1 through 11,14., $nd 16)
The recommendations address specific actions or cautions that should be incorporated in the operating procedures for steah; generator tube rupture elents.
The ProcedLres and Test Review Branch (PTRB) has a program to imp *JV6 opera +.ing procedures through a systematic, analytic approach that will
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deal with emergency procedures for all transients and accidents including those for a steam generator tube rupture. The present program and schedule for developing ingroved emergency operating procedures (E0Ps) for operating reactors is as fMlows:
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The Generic Technical Guidelines for emerg.;ecy operating procedures were originally scheduled to be completed b'y Jairaary 1981. However, the program ht' required more work than ditticipated and the schedule was revised to ria-1982. The staff has net wita the vendors and owners groups over the past year to review and comment ci drafts.
This item has been combhed with other items in the d,af t " Emergency Response Capability and Facilities" prepared by the Canmittee for
. the Re.
of Generic Requirements and is now being considered by that Committee, 2.
URC staff review and approval of the four vendor technical guidelines two months after their submittal. This is considered an achievable schedule, since the task is given a high priority.
3.
Issue industry-prepared writer's guide by June 1982.
This is consistent with INP0's schedule for the task.
4 Issue revised NUREG-0799, 9uidelines for Preparation of Emergency Operating Procedures, to incorporate public comments by June 1982. A draft report was issued for public comment on June 1981.
We consider this program,in the sequence indicated, to be the most effective alternative realistically available.
It is consistent with the ongoing program for improving E0Ps.
The operator training would be most effective if based on the improved procedures. The program provides an integrated evaluation of plant operation which should minimize conflicting operating requirements.
INP0 is supportive of our schedule for completing the guidance documents and is in agreement with our desire to keep the procedures improvement program moving.
B.
Radiation Monitoring Equipment (Recommendation for licensees #12) r, recommendation addresses radiation monitoring equipment of all re# ease points: condenser steam jet air ejectors (SJAEs), steam
. generator atmospheric dump valves (ADVs), auxiliary feedwater (AFW) turbine exhaust and steam generator blowdown lines. The intent of this recommendation is covered by TMI Action Plan II.F.1.
(NUREG-0737, pp. 3-95 ff.) This Action Plan item calls for high range monitors on either the tailpipes of the steam generator safety valves, atmospheric dump valves, and auxiliary feedwater turbines or on the main steam pipes upstream of the takeoff points for these devices.
(High range mnitoring of the steam jet air ejectors at the condenser is also required.) Thus, the radioactive discharge pathways listed in Recommendation 12 will be covered when the Action Plan item is implemented.
However, these monitors will not necessarily discriminate between steam generators as was also intended by the recommendation. Although it is obviously desirable to have individual detectors associated with individual steam lines, and thus be able to discriminate in this way which steam generator is leaking, there are, in some cases, significant engineering problems involved.
Several schemes involving detector shields or continuous sampling techniques are being investigated by the industry, but it is evident that some time will be needed for develcpment. Meanwhile some plants are installing monitors that car.not discriminate between steam generators.
Originally, the monitoring equipment described in Action Plan II.F.1 had an installation deadline of January 1,1982.
Delays due to procurement problems have occurred and only 16 plants have met the
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deadline.
New commitment dates have been set for the remainder of the operating plants, and all but three plants are scheduled to finish installation in 1982.
C.
Reactor Coolant Radioactivity Limits (Recommendation for licensees #13)
The recommendation is that a limit on reactor coolant activity be incorporated in the Technical Specifications of the plants that do not have such a limit.
All plants have a limit on ectivity in the coolant.
liowever,11 of the older PWR plar.ts at eight sites
- have limits only on total activity, which because of the overriding effect of noble gases, does not provide a meaningful limit on radioactive iodine. Thus, for these plants, the analysis of the thyroid dose as required by Part 100 must be based on an assumed va sue of iodine in the coolant rather than an enforceable limit.
D.
Atmospheric Dump Valve Position Indication (Recommendation for licensees #15)
This recommendation calls for means for determining whether steam generator or steam line atmospheric dump valves are stuck open or leaking.
This proposed requirement is included in Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident", which would require either position or leak flow indication Haddam Neck; Indian Point 2; Ke;;aunee; Point Beach 2; Oconee 1, 2 & 3; Rancho Seco; TMI-1; Zion 1 & 2.
on all secondary side safety and relief valves.
Implementation of Regulatory Guide 1.97 is currently under active review by the Committee for Review of Generic Requirements.
E.
Adequacy of Operator Training on Other Than Site-Specific Simulators (Recommendation for NRC 41)
As part of the DHFS program to develop a staff position on plant specific simulators, non-plant specific simulator training and examination procedures are being evaluated.
RES is also investigating the usage of non-plant specific simulators as an ancilliary effort to their program on training.
Presently, the Operator Licensing Branch is administering simulator examina-tions, per TMI Action Plan Item I. A.3.1, on both types of simulators. A Commission paper is being drafted which evaluates the benefits and costs of continuing non-plant specific simulator examinations.
F.
Review of System Performance (Recommendation for NRC #2)
The intent of the recommendation is to improve t'.e analysis and description of steam generator tube rupture events.
The standard Re-view Plan (NUREG-75/087, SRP 15.6.3) previously concentrated on the radiological consequences of a steam generator tube rupture event.
The revised Standard Review Plan (NUREG-0800, SRP 15.6.3, Rev. 2 - July 1981) now calls for RSB review of the sequence of events and plant pro-cedures for recovery. Thus, this recommendation has been adopted.
G.
Main Steamiine Isolation Valve Closure Logic (Recommendation for NRC #3)
This recommendation highlights the possible deficiency in the review of steam generator tube rupture events, i.e., the effect of the main steamline isolation valve closure logic in the event.This is already called for in the Standard Review Plan (NUREG-0800, Rev. 2, p.15.6.3-
- 3) which requires a full review of the applicant's description of events.
H.
Tube Cracking ~
(Recommendation for NRC #4)
The recommendation is to continue to study the phenomenon associated with stress corrosion cracking and implement measures to eliminate it.
The NRC steam generator research program includes a program started in FY 1978 to develop models to predict the stress-corrosion cracking service life of Inconel 600 tubes. The results of this program to date are presented in the Brookhaven National Laboratory, Water Reactor Safety Research Division, February 1981 Quarterly Progress Report, NUREG/CR-1960.
This program is budgeted through FY 1983.
The Connission was recently briefed on the status of the issue of steam generator tube integrity on December 4, 1981 and presented in SECY 81-664, November 24, 1981. The report on the " Resolution of Unresolved Safety Issue A-3, A-4 and A-5 Regarding Steam Generatcr Tube Integrity".
NUREG-0844 is now in the final stages of preparation. The current requirements for steam generator tube surveillance which are incorporated in the plant Technical Specifications are contained in Regulatory Guide 1.83, Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes. The current criteria for tube plugging are coatained in
. Regulatory Guide 1.121 " Bases for Plugging Degraded PWR Steam Generator Tubes".
I.
Undesirable System Conditions (Recommendation for NRC #5)
This recommendation is identical to Applicant Recommendation 16, except that it urges the staff to encourage Westinghouse ta ensure that applicants follow Recommendation 16.
J.
Inspection TRecommendation for NRc #6)
This recommendation addresses assurance that the other requirements are implemented.
IE inspectors automatically monitor compliance with require-ments once they are imposed. The fact that most of the recommendations given here are being addressed by other programs makes it impractical to specifically call out steam generator tube rupture requirements.
II.
Recommendations in "Calvert Cliffs Unit 1 Loss of Service Water on May 20,1980", AE00, December 1981.
A, Loss of Instrument Air T Mcommendation b)
These recommendations are directed toward improving plant procedures and equipment to control the consequences of the rupture of steam generator tubes concurrent with the loss of compressed air and thus air operated equipment, such as atmospheric dump valves, and turbine bypass valves.
NRR is now reviewing the AE0D report and expects to provide a response to the recommendations in the next few weeks.
I NUREG-0651 i
X..-__.-.___-.
l Evaluation of Steam Generator l
Tube Rupture Events I
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Manuscript Completed: March 1980 Date Published: March 1980 L. B. Marsh Division of Operating Reactors Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission i
Washington, D.C. 20555 p m.,
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3.
STAFF RECOMMENDATIONS The three SGT rupture accidents, while having generally acceptable consequence with respect to offsite doses, had the potential for more significant conse-quences.
To ensure that any subsequent SGT rupture events do not result in unacceptable results, the staf f recommends that every Westinghouse PWR licensee implement the following:
(1) Licensees should investigate other techniques to more rapidly determine the damaged SG, and these techniques should be included in the procedures and operator training.
One such technique may be to install radiation monitors in separate steam supply lines from each SG to an automatically starting AFW turbine.*
(2) The timely depressurization of the plant should be emphasized in the plant procedures and in operator training.
Every licensee should adequately emphasize this important facet of the SGT rupture incident.
In addition, RCS subcooling should be emphasized in plant procedures and operator l
training.
(3) Licensees should ensure that their procedures require timely securing of all feedwater to the damaged SC as soon as it is identified and isolated.
The procedures may allow intermittent feeding of the SG should its level mandate such action.
I (4) All licensees should ensure that their procedures and operator training program emphasize the need to expeditiously secure steam flow from the damaged SG to the TD AFP.
The TD AFP should not be started (manually) unless the damaged SG has been identified, isolated, and steam from that
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SG to the TD AFP has been isolated.
If the TD AFP has been auto":atically started, then it should be secured if the other AFPs are operating and adequate feed flow exists.
If the steam to the TD AFP is knwn to be i
from an undamaged SG, then there should be no significant releases in the turbine exhaust and, therefore, no reason to secure the TD AFP.
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Procedural cautions and guidance as to when to secure the TD AFP, if not needed for safe plant operation, should be noted.
As a minimum, running i
times should be logged to later determine radioactive release amounts.
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(5) PWR licensees should implement adequate procedures and training to ensure i
the operator's cognizance of a possible MSIV closure during the r aldown following an SIS.
(The staff should continue to study this feature to ensure its design adequacy with respect to SGT rupture and steam line
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break events.) The operator should be given corrective actions to implement immediately.
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Nestinghouse has suggested that radiation monitors for each main steam line, coupled with appropriate fiow instruction, should be considered.
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(6) The emergency procedure should direct the operator to attempt control of the SGT rupture by the use of charging and letdown systems and to initiate an orderly plant shutdown, if possible.
Reactor trips from high power conditions could easily lead to lifting of SG safety valves and/or ADVs and result in direct radioactive release to the environment.
If the decrease in pressurizer level and pressure is not controllable, then a manual reactor trip should be initiated before the indicated pressurizer level goes of fscale (low).
(7)
If the RCPs are not tripped manually as required by NUREG-0623, the RCP in the damaged loop should not be tripped.
The reduction in RCS spray flow is not significant, but judging from the events being evaluated, RCS depressurization following the event has been slow, and the operator should have as much spray flow available as possible.
Also, heat transfer from the damaged SG to the RCS, while small, is enhanced by greater RCS flow.
Therefore, an operating RCP in the affected loop would aid in the cooldown of the damaged SG.*
(8) For those plants provided with loop isolation valves, the use of these valves following an SGT rupture should be investigated.
Isolating the affected loop would provide an almost immediate abatement of SGT leakage, but would prohibit cooldown of the damaged SG.
Licensees should, therefore, examine the advantages and disadvantages in their plant of loop isolation.
(9) Following an SGT rupture event, SI actuation may occur on low pressurizer pressure.
If RCS pressure continues to fall below the plant-specific predetermined value described in NUREG-0623, the RCPs should be tripped.
(10) Licerisees should develop procedures to restart the RCPs af ter an SGT rupture event if they were tripped as a result of low pressure SI actuation concurrent with RCS pressure falling below a predetermined, plant-specific pressure (as specified in NUREG-0623).
The procedures should, as a minimura, require positive diagnosis of an SGT rupture accident, adequate RCS subcooling, adequate pressurizer level, and the fulfillment of all other RCP restart requirements.
(11) The use of the pressurizer PORV may be necessary following an SGT rupture accident.
If the PORV is required to control RCS pressure upon loss of normal spray capability, guidance shouid be provided to the operator to monitor and control (if possible) the pressurizer. relief tank parameters to minimize the potential of rupture disk relief to containment.
(12) Licensees should review and upgrade radiation monitoring equipment (and its associated surveillance requirements) to assure that reliable monitoring 1
of all release points (i.e., steam jet air ejector, ADVs, AFW turbine, and SG blowdown lines) is available during an SGT rupture accident.
In addition, the operators should be properly instructed in the interpreta-tion of valuable diagnostic information available from a properly functioning RMS.
The system can contribute to an early diagnosis of the accident, as well as to a reliable identification of the f aulted SG.
- Westinghouse disagrees with the staff and states that the RCP in the damaged loop should be tripped to enhance RCS subcooling.
The staff is presently evaluating this assertion with respect to RCS subcooling.
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(13) The limits on coolant activity given in the Standard Technical Specifications should be implemented at the remaining PWR plants without these limiting conditions for operation on a priority basis to ensure that the radiological consequences of SGT rupture accidents will not exceed a small fraction of the 10 CFR 100 guidelines.
(14) To assure correct operator response and plant controllability when bulletin requirements affect emergency procedures, the procedural changes and the operator response should be evaluated using the simulators.
(15) Licensees should investigate the means available to the plant operators to determine whether an SG or steam line ADV is either stuck open or leaking.
Because the inadvertent or improper operation of these valves j
can result in a significant increase in the offsite doses during an SGT rupture, licensees should implement measures to positively identify and isolate an improperly operating ADV on the damaged SG.
If appropriate, upgrading of the maintenance and surveillance requirements on these valves should be incorporated.
(16) Westinghouse has stated that it is possible for one or more of the following situations to occur during the SGT rupture accident (either before or after operator intervention):
(3) Automatic opening of the pressurizer PORV(s) and/or the SV(s)
(b) Water soiid or drained pressurizer (c) Saturation conditions in the RCS Every licensee must ensure adequate procedural steps and operator training so that proper identification and correction of the condition can be performed.
(See item (5) below.)
As a result of our review of the three SGT rupture accidents, the following staff actions are also recommended:
(1) A sample audit of operator training using site procedures on other than site-specific simulators should be conducted by NRC to determine whether the training received is realistic and practical.
A determination should be made as to whether this training is adequate for the operator to relate the emergency procedures to his specific control boards and plant systems.
(2) Future reviews of SGT Tture accident analyses for construction phase (CP) and operating license (OL) stage plants should require a more detailed description of the system performance during the event.
Plots of the various parameters should be provided so that a better understanding and documentation of the event can be achieved.
(3) The staff, together with two-loop Westinghouse PWR licensees, should investi-gate the MSIV closure logic to ensure that its operation does not violate any assumptions in the respective FSAR analyses.
The staff should also look at the FSAR analyses where MSIV operation is assumed, and assure that the logic is consistent with the required operation.
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(4) The staff, together with Westinghouse, should continue to study the phenomena associated with stress corrosion cracking in the U-bend region of the Row 1 tubes and implement measures necessary to eliminate Row 1 tube cracking.
(5) The s+.af f, together with Westinghouse, should determine the potential for u.e undesirable system conditions listed previously in item (16) during an SGT rupture accident and the subsequent RCS cooldown/depressurization for each PWR.
Since the system conditions will depend on SGT rupture size, SIS design characteristics, and specific operator actions, the staff, together with Westinghouse, must ensure that the analyses adequately cover the different plant designs and situations.
(6) The NRC I&E inspectors should enst e implementation of the recommendations for Westinghouse PWR licensees stated earlier in this section and should review and ensure licensees' conformance with these items.
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