ML20195G230
| ML20195G230 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/02/1987 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Jordan E NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| Shared Package | |
| ML20154D718 | List: |
| References | |
| IEIN-87-023, IEIN-87-23, NUDOCS 8706180368 | |
| Download: ML20195G230 (14) | |
Text
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'4 UNITED S TATEs
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NUCLEAR REGULATORY COMMISSION j
WASHINGTON, D, C. 20555 e
June 2, 1987 MEMORANDUM FOR:
Edward Jordan, Director Office of Analysis and Evaluation of Operational Data FROM:
Thomas E. Murley, Director Office of Nuclear Reactor Regulation
SUBJECT:
NRR PLANS FOR RESPONSE TO LESSONS LEARNED FROM DIABLO CANYON LOSS OF RHR EVENT OF APRIL 10, 1987 AND RELATED EVENTS Diablo Canyon Unit 2, a Westinghouse four loop PWR, experienced a loss of RHR on April 10, 1987 that continued for 85 minutes. The core heated to boiling within 30 to 45 minutes, and for the remainder of the event was cooled by reflux condensation in the steam generators.
An Augmented Inspection Team (AIT) was dispatched to the site, and spent more than a vack conducting the onsite investigation. The AIT concluded that lessons learned from the event are of significance to safety and many of the lessons appear applicable to all PWRs.
The staff considered these lessons to be sufficiently important that an Information Notice was issued (87-23, attacned), and individual plants have been contacted if they were in mid-loop or anticipating mid-loop operation.
Two licensees reacted to the lessons learned by voluntarily closing containment a'nd restricting other operations while in mid-loop operation. We have talked with other plant personnel who are reviewing their procedures and hardware in light of the lessons learned information.
Industry contacts regarding this issue are ongo';ng.
The NRR staff has prepared a 50.54(f) letter (attached) which provides additional infonnation, and which requires information from all PWR licensees pertaining to this issue. This letter has been coordinated with and reviewed by AE00, RES, OGC, and the Augmented Inspection Team leader. The AIT report is expected to be available by mid-June 1987. OGC has reviewed this package and has no legal objection.
Contact:
W. Lyon, SRXB, x27605 i
a 4
1%lVG!b hy,
Edward Jordan We request a CRGR review of this 50.54(f) letter be scheduled at the earliest opportunity.
,$ AM
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s E. Murley, Director Of ice of Nuclear Reactor Regulation
Enclosures:
As stated cc:
V. Stello J. Grace A. Davis R. Martin J. Martin l
i i
1 ATTACHMENT 1 RESFONSE TO REQUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW 1.
A problem statement that describes the need for the infonnation in terms of potential safety benefit:
The Diablo Canyon loss of RHR event during mid-loop operation could reasonably have resulted in unanalyzed overpressurization of the RCS with a potential for fuel damage and possibly more severe consequences under reduced RCS inventory conditions. A detailed description of the conditions during mid-loop operation--information such as instrurnentation requirements at the plant, procedural requirements, operator training, and boundary conditions (like decay heat loads, for example)--is necessary to assure safe operation in this mode in accordance with the licensing bases and GDC-34.
2.
The licensee actions required and the cost to develop a response to tne information request:
The licensee would be expected to supply detailed descriptions of plant RCS conditions during mid-loop operation, procedures required during ais mode cf operation with any restrictions and analytical bases that may apply, information pertaining to training of personnel involved in this mode of operation, and proper installation of available equipment. The cost of providing this infomation would be minimal.
3.
An anticipated schedule for NRC use of the information:
Assuming 60 days is given to the licensees to respond to the twuest for infonnation, we estimate the review to be complete by April 30, 1988.
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ATTACHMENT 2 TO:
All Licensees of operating PWRs and holders of construction pennits for PWRs Gentlemen:
SUBJECT:
LOSS OF RESIDUAL HEAT REMOYAL (RHR) DURING MID-LOOP OPERATION
- Pursuant to 10 CFR 50.54(f), the NRC is requesting infonnation to assess safe operation of Pressurized Water Reactors (PWRs) when the Reactor Coolant System (RCS) water level is below the top cf the Reactor Yessel (RV). The principal concerns are (1) failure of the RHR system to meet the design basis of the plant, such as General Design Criterion 34 (10 CFR Part 50, Appendix. A), and Technical Specifications (TS), in this condition; and (2) the resultant unana-lyzed impact upon safety.
Our concerns regarding this issue have increased over the past several years, and lessons learned from the April 10, 1987 Diablo Canyon loss of RHR event require an assessment of operations and planned operations at all PWR facili-ties to insure that these plants meet this licensing basis. Study of the Diablo Canyon event has led to identification of unanalyzed conditions which are of significance to safety. Although Diablo Canyon never came close to core damage, and could have withstood the loss of RHR condition for over a day with no operator action, slightly different conditions could have led to a core damage accident within several hours. One unanalyzed condition involves boiling within the RCS in the presence of air, leading to RCS pressurization with the potential for ejecting RCS water via cold leg openings, such as could exist during Reactor Coolant Pump (RCP) or loop isolation valve repairs.
The lost water would no longer be available to cool the core, and this could significantly decrease the time to core damage if makeup were unavailable.
The pressurization could also affect the capability to provide makeup water to the core. Other unanalyzed situations are also possible, and occurred at Diablo Canyon (e.g., boiling in the core). The seriousness of this situation is exacerbated by the practice of conducting operations with the equipment hatch recoved, and by the lack of procedures which address prompt containment isola-tion should the need arise.
Loss of RHR and related topics are not a new concern to the NRC staff. This topic has been addressed in numerous comunications with the licenses. Yet, events continue to occur at a rate of several per year. This condition needs to be fully considered in order to ensure compliance with Cennission require-ments. Therefore, we request you to conduct an safety assessment of operation of your plant during the approach to mid-loop condition and while in that operating condition to assure that you meet the licensing basis. Your safety assessment is to include the following:
Shid-loop operation as used here is the condition where water level in the reactor coolant system is below the level of the top of the reactor vessel.
, 1.
A detailed description of the circumstances and conditions under which your plant would be entered into and brought through a draindown process and operated at mid-loop conditions.
Examples of the type of infonnation required are the time between full power operation and reaching a mid-loop condition used to generate values for decay heat loads, requirements for minimum SG levels, restrictions regarding removal of equipment for maintenance and testing while in mid-loop, restrictions regarding testing and maintenance that could perturb the NSSS, requirements pertaining to isolation of containment, and the time required to replace the equipment hatch should replacement be necessary.
2.
A detailed description of the instrumentation and alanns provided to the operators for control of thermal and hydraulic aspects of the NSSS during operation in mid-loop operation.
You should deuribe temporary piping used for instrumentation and the quality control prccess to assure proper functioning of such connections, piping, and instrumentation, including assurance that they do not contribute to loss of RCS inventory or othemise lead to perturbation of the NSSS during mid-loop operation.
3.
Identification of all punps which can be used for centrol of NSSS inventory.
Include:
a.
Pumps you require be operable or capable of operation. Where such pumps may be temporarily removed from service for testing or maintenance, such information is to be included, b.
Other pumps not included in "a".
c.
An evaluation of the above with respect to applicable Technical Specification (TS) requirements.
4.
A description of the containment closure condition you require for the conduct of operations during mid-loop operations.
Examples of areas of consideration are the equipment hatch, personnel batches, containment purge valves, SG secondary side condition upst; sm of the isolation valves (including the valves), piping penetraticns, and electric)1 penetrations.
5.
Reference to and a sumary description of procedures in the control room of your plant which describe operation for mid-loop operation. Your response should include the analysis basis used for procedures developnent, including a description of analyses which illustrate Nuclear Steam Supply Sy5 tem (NSSS) response to norinal operation and to mitigative actions. We are particularly interested in your treatment of drain @wn to mid-loop, analysis of minor variations from expected behavior such as due to air entraircEnt and de-entrainment, boiling in the core with and without RCS pressure boundary integrity, calculations of approdnste tilte to core damage, vortexing, level differences in the RCS and t1he effect upon instrumentation indications, and treatment of air in the RCS/RHR system, including the impact of air upon NSSS and instrumentation response.
The analysis should support the following:
3 a.
Procedural guidance pertinent to timing of operations, required instrumentation, cautions and critical parameters, b.
Operations control and communications requirements regarding opera-tions which may perturb the NSSS, including restrictions upon testing and maintenance operations which could upset the condition of the NSSS.
c.
Response to loss of RHR, including regaining control of RCS heat removal, operations involving the NSSS if RHR cannot be restored, control of effluent from the containment if containment was not in an isolated condition at the time of loss of RHR, and operations to provide containment isolation if containment was not isolated at the time of loss of RHR. Guidance pertinent to timing of operations, cautions and warnings, critical parameters, and notifications is to be clearly described.
6.
A brief description of training provided to operators and other affected personnel that is specific to the issue of mid-loop operation. We are particularly interested in such areas as maintenance personnel training regarding avoidance of perturbing the NSSS and response to loss of decay heat removal during mid-loop operation.
7.
Identification of additional resources provided to the operators during mid-loop operation, such as assignment of additional personnel with specialized knowledge involving the phenomena and instrumentation.
8.
Comparison of the requirements implemented for mid-loop operation and requirements used in other Mode 5 operations.
Some requirements and procedures followed during mid-loop operation may not appear in the other modes. An example of such differences is the operation with a reduced RHR flow rate to minimize the likelihood of vortexing and air ingestion.
9.
As a result of your consideration of these issues, you may have made changes to your current program related to these issues.
If such changes have strengthened your ability to operate safely during a drained-down situation, then please provide descriptions of those changes and scheduling information.
' contains insight which experience indicates should be well-understood prior to comencing mid-loop cperation. Additional information will be contained in the NRC Augmented Inspection Team report, NUREG 1269, "Loss of Residual Heat Removal System Diablo Canyon Unit 2 April 10,1987", a draf t copy of which will be forwarded to you in the near future.
Your response addressing items 1 thru 9 above is to be signed under oath or affirmation, as specified in 10 CFR 50.54(f), and will be used to detennine whether or not your license should be modified, suspended, or revoked. We request your response within 60 days of receipt of this letter. This information is required to assess conformance of PWRs with their licensing
4 basis and is therefore exempt from backfit requirements. Our review of your submittal of information is rot subject to fees under the provision of 10 CFR 170. We suggest you consider providing a portion of your response in association with your respective owners group since much of this issue is of generic origin.
This request for information was approved by the Office of Management and Budget under clearance number 3150-0011 which expires September 30, 1989.
Comments on burden and duplication r,ay be directed to the Office of Management and Budget. Reports Management Rcom 3208, New Executive Office Building, Washington D. C. 20503.
Sincerely, l
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a ENCLOSURE 1 TO ATTACHMENT 2 INFORMATION PERTINENT TO LOSS OF RESIDUAL HEAT REMOVAL SYSTEMS WHILE IN MID-LOOP OPERATION Many maintenance and test activities conduct 2d during an outage require lowering the Reactor Coolant System (RCS) water level to below the top of the Reactor Vessel (RV) and many times to the centerline elevation of the RV nozzles. This operating regime is typically known as "mid-loop" operation.
It places un9sual demands upon plant equipment and operators due to narrow control margins and limitations associated with equipment, instrumentation, i
procedures, training, and the ability to isolate containment.
Difficulty in l
l controlling the plant while in this condition often leads to loss of the l
Residual Heat Removal (RHR) System, as illustrated in' Table 1.
Although thit issue has been the topic of many comunications and investigations, events continue to occur at a rate of several per year.
Recent knowledge has provided additional insight into these events. Although the full implications of this knowledge remain to be realized, our preliminary assessments have clearly established real and potential inadequacies associated with mid-loop operation. These include not understanding the Nuclear Steam Supply System (NSSS) response to loss of RHR, inadequate instrurentation, lack of analyses which address the issue, lack of applicable procedures and training, and failure to adequately address the safety impact of loss of decay heat removal capability.
The following items are applicable to these conclusions:
1.
Plants enter an unanalyzed condition if boiling occurs following loss of RHR. For exarple:
a.
U,iexpected RCS pressurization can occur.
2 No pressurization would occur with a water / steam filled RCS with water on the Steam Generator (SG) secondary side as RCS steam would condense in the SG tubes and the condensate would return to the RV.
Air in the RCS can block the flow of steam through passages, such as the entrance portion of SG tubes, so that steam cannot reach cool surfaces.
Failure to condense the steam causes RCS pressurization until sufficient compression of the air occurs that steam can reach cooled tube surfaces. This pressurization occurred during the April 10, 1987 event at Diablo Canyon since the RCS contained air.
Pressure reached 5 to 10 psig, and would have continued to increase if RHR had not been restored.
The operators initiated event tennination by allowing water to flow from the Refueling Water Storage Tank (RWST) into the RCS.
Increasing pressure would have eliminated this option, and would have jeopardized options involving pumps with suction lines aligned (in part) to the RCS, b.
Water that ordinarily would be available to cool the core might be forced out of the RV, thereby reducing the time between loss of RHR and initiation of core damage.
This is a potential concern whenever there is an opening in the cold leg, such as may exist for repair of Reactor Coolant Pumps (RCPs) or loop isolation valves.
Upper vessel / hot leg pressurization could t
force the RV water level down with the displaced water lost through the cold leg opening. A corresponding level decrease would occur in the SG side of the crrssover pipes between the SGs and the RCPs.
This occurrence could be particularly serious if the cold leg opening were large or makeup flow to the RCS small, as from a charging pump. Cold leg injection with elevated pressure in the upper vessel may not provide water to the core. Hot leg injection would probably be effective.
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3 2.
RCS water level instrumentation may provide inaccurate infonnation.
There are many facets to this issue.
Instrumentation may be indicating a level that differs from level at the RHR suction line, a temporary instrument may be in use with no indication or alanns in the control room, and design and installation deficiencies may exist. We have observed the following:
a.
Connections to the RCS actually provide a water level indication up-stream of the RCP location.
This water level is higher than the water level at the RHR suction connection due to flow from the injection to the suction locations and due to entering water acmentum, which increases level on the RCP side of the cold leg injection location.
Ingestion of air at the RHR suction connection will result in transporting air into the cold legs, which can potentially increase pressure in the air space in the cold legs relative to the hot legs.
Level instrumentation may respond to such a pressure change as though RCS level were changing.
In addition, such a pressurization would move cold leg water into the hot legs and upper RV (or the reverse if a depressurization occurs),
b.
Use of leng, small diameter tubing which can lengthen instrument response time and cause perturbations such as RCS pressure changes to appear as level changes, tubing elevation changes which can trap air bubbles or water droplets, and tubing which can become kinked or constricted, c.
Some installations provide no indication in the control room, yet level is important to safety. Some provide one indication. Others provide diversity via different instrumentation, but do not provide independence due to coninon connections.
4 d.
Tygon tube installations with faint level marks at one foot intervals, with no provision for holding the tube in place.
e.
Instrumentation in which critical inspections were not perfonned after the installation.
f.
Instrumentation where no provisions were made to assure a single phase in connection tubing or that tubing was not plugged.
g.
Use of instrumentation without performing an evaluation of indicated RCS level behavior and instrument response.
3.
Vortexing and air ingestion from the RCS into the RHR suction line are not always understood, nor is NSSS response understood for this condition, a.
On April 10, 1987, Diablo Canyon operators reduced indicated RCS level to plant elevation 106' 6" imediately following SG tube draining, and observed erratic RHR pump current indications.
Restoration of level to 106' 10" was reported to have eliminated the problem. RHR operation wat terminated a few hours later at an indicated level of 107' 4" due to observed erratic RHR pump current indication. The Licensee later reported that vortexing initiated under those conditions at 107' S 1/2", and was fully developed at t
107' 3 1/2". Procedures in place at the time of the event indicated the minimum allowable level to be 197' 0" (the hot and cold leg centerline elevation) or 107' 3".
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b.
Additional phenomena appear to occur under air ingestion conditions, i
These include:
(1) RHR pumps at Diablo Canyon were reported to handle several 3
l percent air with no discernible flow or pump current change from that of single phase operation.
I
5 (2) One rostulate is that air in the RHR/ Reactor Coolant system can migrate or redistribute, and thus cause level changes which are at variance with those one would expect. This is a possible explanation for observed behavior where a lowering of RCS water level is followed by a level increase. Water in the RHR appears to be replaced by air.
Similarly, an increase in RCS water level that is followed by a decreasing level may be due to voids in the pHR system being replaced by RCS water.
Failure to understand such behavior leads to operator mistrust of level instrumentation and to operational errors.
c.
Operators typically will s; art another RHR pump if the operating purp is lost. Experience and an understan/,ing of the phenomena clearly show that loss of the second pump should be expected. The cause of loss of the first pump should be well-understood and normally be corrected before attempting to run another RHR pump, d.
Typical mid-loop operation provides a high RHR flow rate, which may be required by TS, but which may be unnecessary under the unique conditions associated with mid-loop operation. Air ingestion problems are less at low flow rates.
4.
Only limited instrumentation may be available to the operator while in mid-loop operation, a.
Level indication is many times available only in containment via a Tygon tube.
Some plants provide one or more level indications in the control room, and additionally provide level alarms, b.
Typically, RHR system terrperature indication is the only temperature provided to the operators. Loss of RHR leaves the operator with no RCS teftperature indication. This can result in violation of Technical Specifications, as occurred at Diablo Canyon on April 10
6 when the plant entered Mode 4. unknown to the operators, with the containment equipment hatch removed.
It also resulted in failure to recognize the seriousness of the heatup rate, or that boiling had initiated.
c.
RHR pump motor current and flow rate may not be alarmed and scales may not be suitable for mid-loop operation.
I d.
RHR suction and discharge pressures may not be alarmed and scales may not be suitable for mid-loop operation.
5.
Licensees typically conduct mid-loop operations with the containment equipment hatch removed and with operations in progress which impact the ability to isolate containment.
Planning, procedures, and training do not address containment closure in response to loss of RHR or core damage events. This is inconsistent with the sensitivity associated with mid-loop operation and the history of loss of RHR under this operating I
condition.
l 6.
Licensees typically conduct test and maintenance operations which can
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perturb the RCS and RHR system while in mid-loop operation. The sensitivity of mid-loop operation and the historical record indicate this is not a prudent activity.
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. o ATTACHMENT 3 SSINS No.: 6835 IN 87-23 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555 May 27, 1987 s
NPC INFORMATION NOTICE NO. 87-23: LOSS OF DECAY HEAT REMOVAL DURING LOW REACTOR COOLANT LEVEL OPEPATION o
Addressees:
All holders of an operating license or a construction permit for pressurized-water rea: tor facilities.
Purpose:
This notice provides infonnation reoarding the loss of decay heat removal capability at pressurized water reactors resulting from the loss of PPR pump suction during plant operations with low reactor coolant levels. It is ex-pected that recipients will review this infonnation for applicability to their reactor facilities and consider actions, if appropriate, to prevent similar problems. Suggestions contained in this notice do not constitute NRC require-ments; therefore, no specific action or written response is required.
Description of Cireurstances:
On April 10, 1987 the Diablo Canyon Unit 2 reactor experienced a loss of decay heat removal capability in both trains.
The reactor coolant system had been drained down to the mid-height of the hot-leg piping in preparation for the rer. oval of the steam generator manways.
During the 85 minute period that the heat-removal capability was lost, the reactor coolant heated from 87' F to boiling, steam was vented from an opening in the head, water was spilled from the partially unsealed manways, and the airborne radioactivity levels in the l
containment rose above the maximum pennissible concentration of noble gases allowed by 10 CFR 20.
The reactor, which was undergoing its first refueling, i
had been shut down for seven days at the time and the containment equipment hatch had been opened.
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Erroneous level instrumentation, inadequate knowledge of pump suction head / flow requirteents, incerplete assessment of the behavior of the air / water mixture in i
the system and poor coordination between control room operations and contain-l ment activities all contributed to the event. Under the conditions that l
existed, the system that indicated the level of coolant in the reactor vessel read "high" and responded pocrly to changes in the coolant level.
In addition, the intended coolant level, established for this operation, was later deter-mined to be below the level at which air entrainment due to vortexing was predicted to comence. At the time of the event, the plant staff believed that the coolant level was six inches or more above the level that would allow vortexing.
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- C.700749
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IN 87-23 Pay 27, 1987 Page 2 of 5 The event began at about 8:43 pm, when a test engineer in preparation for a planned containment penetration local leak rate test, begar draining a section of the reactor coolant pump leakoff return line, which he believed to be iselated. However, because of a leaking boundary valve, this acti9n caused the volume cor. trol tank fluid to be drained through the interded test section to the reactor coolant drain tank. The control room operators, who were not aware that the engineer had begun conducting the test procedure, increased flow to stop the fluid reduction from the volume control tank. A few minutes later the operators were informed that the reactor coolant drain tank level was increas-ing bul they could not determine the, source of the leakage. Although the actual level of coolant in the reactor vessel was apparently dropping below the minimum intended level, the indication of level in the vessel remained within the desired control band. At 9:25 p.m. the electrical current of the active RHR pump (No. 2-2) was observed to be fluctuating. The 2-1 pump was started and the 2-2 pump was shut down. However, the current on the 2-1 pump also fluctuated, so it was inrediately shut down as well.
The operators did not imediately raise the water level in the reactor because they still did not know either the source of the leekage, the true vessel level, or the status of the work on the steam generator manways. Operators were sent to vent the RHP pumps. One pump was reported to be vented at 10:03 At 10:21 p.m. an attempt was made to start this RHR pump, but the current p.r.
fluctuated and it was shut down again. During this period the operators did not know the temperature of the coolant in the reactor vessel because the core exit therrocouples had been disconnected in preparation for the planned refuel-ing.
By 10:30 p.m. airborne activity levels in the containment were increasino and personnel began to evacuate from the containment buildiep.
At 10:38 p.m. when the operators learned that the steam generator manways had not been removed, action was initiated to raise the reactor vessel water level by adding water from the refueling water storage tank. About 10 minutes later the test engineer identified the source of the leakage and stopped it. By 10:51 p.m., the vessel level had been raised sufficiently to restart one of the RPP pumps. The indicated RHR pump discharoe teeperature imediately rose to 220' F.
At this tire the reactor vessel was slightly above atmospheric pres-sure and steam was venting from an opening in the reactor vessel head.
Discussion:
The NRC has docunented numerous instances in the past where decay heat removal systems have been disabled because pump suction was lost while the plant was being operated at low reactor coolant water levels.
IE Informatior Notice 86-101 describes four such events that occurred in 1985 and 1986. NRC Case Study Report AE00/C503 describes six such events that occurred in 1984, five that occurred in 1983, and seven that occurred in 1982.
1E Inferination Notice eb09 described an event at Peaver Valley in March 1981. The case study report further indicates that 3 total of 32 such events occurred from 1976 through j
1084 The docunentation includes descriptions of a total of 23 events that have occurred since 1981 involving loss of decay heat renoval capability resulting from a loss of pump suction while operating at reduced water levels.
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IN 87023 May 77, 1987 Page 3 of 5 for all but four of these 23 events the primary cause of the loss of pump suction and loss of decay heat removal capability was attributed to incorrect, inaccurate, or inadequate level indication. Two events were attributed to loss of pump suction because of vortexing brought on by the simultaneous operation of both pumps.
In many of these events procedural errors were also a contrib-uting factor.
In at least nine of the cases, the redundant pump was lost because air was entrained when the operators, not understanding the cause of the problem, switched to the second pump.
There are repeated references to difficulties in getting the pumps vented quickly af ter air binding had occurred and to the operators' inability to tpe imediate action to raise reactor vessel levels until the safety of personnel working on the primary systems could be assured. The length of time that decay heat removal was completely lost varied frnm eight minutes to two hours and averaged almost an hour.
In at least three previous cases, boiling is known to have occurred.
A number of actions have been recomrended previously to prevent the loss of RHR pump suction during low vessel level operations.
These includa:
Providing accurate level instrumentation designed for reduced vessel water level operations.
Providing alares in the control room for low decay heat removal flow and low water level.
Including in the procedures specific requiremnts for frequent monitoring and strict linits on level.
Considering in the procedures the possibility of vortex femation and air entrainment, including a precaution against sta-tin a second RHR pump until the cause of the loss of the first pump is detemined and corrective actions have been taken.
Training the operators on the correlation between water level and pump speed at the onset of vertexing and air entrainment.
Careful planning, coordination, and comunication with control room personnel regarding all ongoing activities which could affect the primary system inventory, i
The NRC review of the Diablo Canyon event indicated that vortexing and air i
entrairment may occur at higher water levels than anticipated.
In addition, operation at mid hot-leg levels can lead to unanticipated conditions which may not have been adequately considered in instrumentation design and procedure l
preparation.
The NRC staff's initial assessment of this event has idertified the potential for a significant loss of decey heat removal capability both from a total loss of the RHR system and from a loss of the steam generator heat sink due to air blanketing of the steam generator tubes.
Correct operator actions then become i
critical for plant recovery.
IN 87-23 May 27, 1987 Page 4 of 5 MPC comunications in the past have expressed serious concern with failures to inaintain adequate decay heat removal capability.
IE Inforration Notice 81-09 pointed out that loss of shutdown cooling capability had been found to be a potentially sionificant contributor to the total risk. AE0D/C503 gnd other sources indicate that the tims available to testore shutdown cooling before core uncovery can occur is not necessarily large. At four days after shutdown frun long-tem power operation, with the vessel draine6 down to the' RHP suction loss level, the vessel water can heat to the boilino point in about 1/? hour.
Under such conditions boiloff to the Aore uncovery level can occur in less than two hours.
Following the loss of decay heat renoval capability on April 10, 1987 at Diablo Canyon, PGIE took a number of actions to prevent loss of RHR suction during low level operation and to improve recovery should such a loss occur. These actio,ns included the following:
Eveluation of the reactor vessel level indicating system to determine the level at which vortexing would occur and the effect of vortexing on the level measurement.
Enhancements of the instrurentation to in:lude accurat? level ireasurement, alam capability and core exit temperature measurerent during low level operation.
Enhancement of procedures to include reovirements fer verifying proper RHR pump suction before starting the second RHR pump. Also included are precautions specifying minitrum vessel levels as a function of PHR flow.
Improvernents in work planning, control and comunication to include a restriction of the work scope to items that do not have the potential to reduce RCS inventory.
Improvement of operator training including a discussion of the potential causes of RHR flow loss, as well as recovery procedures.
The NRC is currently considering additional generic action on this issue.
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IN 87-23 May 27, 1987 Page 5 of 5 This infomation notice requires no specific action or written response.
If you have any cuestions about this matter, please contact the Regional Administrator o' the appropriate regional office or this office.
a d A' c. A-Charles E. Rossi Director Division of Operational Events Assessrent
,' Office of Nuclear Reactor Regulation Techrical Contacts: Donald C Kirkpatrick, NPR (301) 492-8166 Warren C. Lyon, NRR (301)492-7605 j
Attachment:
1.
List of Re:ently Issued NPC Infomation Notices I
}
I
v'ttacnment 1 U. 87-23 May ?7, 10A7 LIST OF RECENTLY ISSUED thFORMAT10N NOTICES 1Q87 Irifoma tion Date of Notice No.
Subiect Issuance issued to a
87-27 Operator Licensing Requali-5/22/87 All research and nonpower. reactor fication Examinations at Nonpower Reactors facilities.
87-21 Shutdown Order issued fecause 5/11/87 All nuclear power Licensed Operators Asleep f acilities holding While on Duty an OL or CP and all licensed operators.
87-20 Hydrogen Leak in Auxiliary 4/20/87 All nuclear power Building facilities bolding an OL or CP 86-108 Degradation of Reactor 4/20/87 All PWQ facilities Sup. 1 Coolant System Pressure holding an OL or CP.
Boundary Resulting from Boric Acid Corrosion 86-64 Deficiencies in Upgrade 4/20/87 All nuclear power Sup. 1 Programs for Plant facilities holding Emergency Operating a CP or OL, Proc *dures.
85-61 Misadministrations to 4/15/8)
All licensees i
Sup. 1 Patients Undergoing Thyroid authorized to use l
Scans byproduct material l
87-19 Perforation and Cracking of 4/9/87 All Westinghouse i
Rod Cluster Control Assemblies power PWR facilities holding an OL or CP 87-18 Unauthorized Service on 4/8/87 All NRC licensees l
Teletherapy Units by Non.
authorized to use licensed Paintenance Personnel radioactive raterial i
in teletherapy units l
l i
87 17 Response Time of Scram A/7/87 All GE BWP facilities 7
instrument Volume level holding an OL or CP Detectors l
l i
OL = Operating License CP = Construction Pemit i
[
l l
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1 l
j Table 1 Chronology of 37 loss of DHR Events Attributed to inadequate RCS Level Docket Plant Date Duration Heatup 344 Trojan 5/21/77 55 min.
Unknown 3/25/78 10 min.
Unknown 3/25/78',
10 min.
Unknown 4/17/78 Unknown Unknown 334 Beaver Valley 1 9/4/78<
60 min.
145 - 175'F 366 Millstone 2 3/4/79 Unknown 150 - 208'F 272 Salem 1 6/30/79 34 min.
Unknown 334 Beaver Valley 1 1/17/80 Unknown Unknown 4/8/80 35 min.
0 4/11/80 70 min.
101 - 108'F 3/5/81 54 min.
102 - 168'F 344 Trojan 6/26/81 75 min.
140 - 150'F 369 McGuire 1 3/2/82 50 min.
105 - 130'F 1
339 North Anna 2 5/20/82 8 min.
Unknown 5/20/82 26 min.
Unknown 5/20/82 60 min.
Unknown 7/30/82 46 min.
Unknown 338 North Anna 1 10/19/82 36 min.
Unknown 10/20/82 33 min.
Unknown 369 McGuire 1 4/5/83 Unknown Unknown 339 North Anna 2 5/3/83 Unknown Unknown 280 Surry 1 5/17/83 Unknown Unknown 328 Sequoyah 2 8/6/83 77 min.
103 - 195'F 370 McGuire 2 12/31/83 43 min.
Unknown 1/9/84 62 min.
Unknown 344 Trojan 5/4/84 40 min.
105 - 201'F 316 OC Cook 2 5/21/84 25 min.
Unknown 368 ANO 2 8/29/84 35 min.
140 - 205'F 295 2 ion 1 9/14/84 45 min.
110 - 147'F 339 North Anna 2 10/le/84 120 min.
Unknown 413 Catawba 1 4/22/85 81 min.
140 - 175'F 327 Sequoyah 1 10/9/85 43 min.
<1'F 296 Zion 2 12/14/85 75 min, s15' 4
361 San Onofre 2 3/26/86 49 min.
114 210'F i
382 Waterford 3 7/14/86 221 min.
138 - 175'F 327 Sequoyah 1 1/28/87 90 min.
95 115'F 323 Diablo Canyon 2 4/10/87 85 min.
100 - 220'F 1
i fm
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