ML20234B434
| ML20234B434 | |
| Person / Time | |
|---|---|
| Issue date: | 06/26/1987 |
| From: | Starostecki R Office of Nuclear Reactor Regulation |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8707060033 | |
| Download: ML20234B434 (19) | |
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JUN 2 6 L MEMORANDUM FOR:
Frank J. Miraglia Associate Director for Projects FROM:
Ricaard W.
Starostecki, Associate Director for Inspection and Technical Assistance
SUBJECT:
TRANSFER OF CRGR APPROVED 10 CFR 50.54(f)
LETTER FOR ISSUANCE TO LICENSEES CONCERNING OIABLO CANYON LOSS-0F-RHR EVENT On Wednesday, June 10, 1987 the Committee to Review Generic Requirements (CRGR) reviewed the NRR proposal to send a 10 CFR 50.54(f) letter to licensees of PWRs.
The letter requires those licensees to provide responses to NRR concerns relating to the Diablo Canyon loss of RHR with the reactor coolant i
loop partially filled.
CRGR voted in favor of the NRR proposal and requested certain modifications to the letter.
' Tis final version of the 10 CFR 50.54(f) letter has been prepared for your issuance.
M TfMI Signerf h K
IL W. SI Arusted,j Richard W. Starostecki, Associate Director for Inspection and Technical Assistance
Enclosures:
As stated
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E. Jordan R. M.
Bernero T. T. Martin D. F. Ross J. Scinto J.
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Sniezek A.
Thomas H.
Smith
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/TTACHMENT 1 TO:
Al' licensees of operating PWRs and holders of construction permits for PWRs Gentlemen:
SUBJECT:
LOSF OF PFSIDUAL HEAT REMOVAL (RHR) WHILE THE REACTOP COOLANT SYSTEM (RCSi IS PAPTIALLY FILLED Pursuant to 10 CFk 50.54(1). the NRC is requesting information to assess safe vperation of pressurized-water reactors (PWRs) when the reactor coolant system (RCS) water level is below the top of the reactor vessel (PV1 The principal l
concerns are (11 whether the RHR system meets the licensing basis of the plant, such as General Design Criterion 34 (10 CFR Part 50, Appendix A) and Technical Specifications (TS), in this condition; (2) whether there is a resultant unanalyzed event that may have an impact upon safety; ind (3) whetner any threat to safety that warrdnts further NRC attention exists in this condition.
Our concerns regarding this issue have increased over the past teveral 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 facilities to ensure that these plants meet the licensing basis.
Study of the Diablo Canyon event has led to identification of unanalyzed conditions that are of significance to safety. Although Diablo Canyon never came close to core ddmage and Could have withstood the loss-of-RHR condition for more than a day s
with no operator action, slightly different conditions could have led to an occident involving core damage within several hours.
One unanalyzed condition involves boilina within the RCS in the preser.ce of air, leading to RCS pressurization with the potential for ejecting RCS water via cold-leg openings, such as could exist during repair to a reactor coolant pump (RCP) or to a loop isolation valve.
The lost water would ro longer be available to cool the core, and if makeup water wert unavailable, the core could be damaged in a significantly decreased time.
The pressurization could also affect the capability to provide makeup water to the core. Other unanalyzed situations dre 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 removed, and by the lack of procedures that address prompt containment isolation should the need arise.
t css of RHR and related topics are not a new concern to the NRC staff. This topic has been addresseo in numerous communications with the licensee.
- Yet, these events continue to occur at a rate of several per year.
This condition needs to be fully considered in order to ensure compliance with the licensing basis, Therefore, we request that you provide the NRC with a description of the operation of your plant during the approach to a partially filled RCS condition and during operation with a partially filled RCS to ensure that you meet the
!nensina basis.
Your description is tu include the following:
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(1) A cetailed description of the circumstances and conditions under which j
t your plant would be entered into and brought through a draindown process I
and operated with the RCE portially filled, including any interlocks that could cause a disturbance to the systeal.
Examples of the type of information required are the time between full-power operation and teaching 6 partially filled condition (used to determine decay heat i
! cads); requirements for minimum steam generator (SGI levels; changes in the status of equiproent for maintenance and testing and coordination of such operations while the RCS is partially filled; restrictions regarding i
testing, operations, and maintenance that could perturb the nuclear steam a
supply system (NSSS); ability of the RCS to withstand pressurization if the reactor vessel head and stean generator manway are in place; requiremerits pertaining to isolation of containment; the time required to replace the equipment hatch should replacement he necessary; and requirements pertinent to reestablishing the integrity of the RCS pressure bou nda ry.
(2) A detailed description of the ir,st rumentation and alarms provided to the operators for controlling thermal and hydraulic aspects of the NSSS during creration with the RCS partially filled.
You should describe temporary cernections, piping, and instrumentation used for this RCS condition and the quality control process to ensure proper functioning of such connections, piping, and instrumentation, including assurance that they do I
not contribute to loss of RCS inventory or otherwise lead to perturbation of the NSSS while the RCS is partially filled.
You should also provide a description of your ability to monitor RCS pressure, temperature, and level af ter the RHR function may be lost.
(3)
Identification of all pumps that can be used to control NSSS inventory, loclude:
(a) pumps ycu require be operable or capable of operation (include information about such pumps that may be temporarily removed from service fnr testing or maintenance); (b) other pumps not included in item a (above); and (c) an evaluation of items a ard b (above) with respect to applicable TS requirements.
(4) A cescriptien of the containment closure condition you require for the conduct of operations while the RCS is partially filled.
Examples of areas of consideration.are the equipment hatch, personnel hatches, containment purge valves, SG secondary-side condition upstream of the isolation valves (including the valves), piping penetrations, and electrical penetrations.
3 (L) Reference to and a summary description of procedures in the control room of your plant which describe operation while the RCS is partially filled.
Your response should include the analytic basis you used for procedures development.
We are particularly interested in your treatment of uraindown to the condition where the RCS is partially filled, treatment of minor variations from expected behavior such as caused by air entrainment and de-entrainment, treatrrent of boiling in the core with and without RCS pressure boundary integrity, calculations of approximate time j
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I from loss of RHR to core damage, level differences in the RCS and the ef fect upon instrumentation indications, treatment of air in the RCS/RHR system, including the impact of cir upon NSSS and instrumentation I
response, and treatment of vortexing at the connection of the RHP suction line(s) to the RCS.
Explain how your an61ytic basis supports the following as pertaining to your facility:
(a) procedural guidance pertinent to timing of operations, required instrumentation, cautions, and critical parameters; (b) operations control and communications requirements regarding operations that may perturb the NSSS, including restrictions upon testing, maintenance, and coordination of operations that could upset the condition of the NSSS; and (cl response to loss of RHR, including regaining control of PCS 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 PHR, and operations to provide containment isniation if containment was not isolated at the time of less 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 trainino provided to operators end other af fected personnel that is specific to the issue of operation while the PCS is partially filled. We are particularly interested in such areas as maintenance personnel training regarding avoidance of perturbing the NS$$
and response to loss of decay heat removal while the RCS is partially filled.
(7)
Identification of additional resources provided to the operators while the RCS is partially filled, such as assignment of additional personnel with specialized knowledge involving the phenomena and instrumentation.
(81 Comparisen of the requirements implemented while the RCS is partially filled and requirements used in other Mode 5 operations. Some requirements and procedures iallowed while the RCS is partially filled may not appear in the other modes. An example of such differences is 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 partially filled situation, describe those changes and tell when they were made or are l
scheduled to be made. contains insight which experience indicates should be well understood before connencing operation with a partially filled RCS. Your response to this 50.54(f) letter request should encompass the topics contained in [nclosure 1.
Add {tional information is contained in the NRC Augmented inspection Team report, NUPFG-1269, " Loss of Residual Heat Removal System, Diatilo Canyon Unit 2, April 10,1987." A copy of NUREG-1269 is enclosed.
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Your responic addressir.g items 1 through 9 (above) is to be signed under oath or affirmation, as specified in 10 CFP 50.54(f), and will be used to determine whether your licence should be niodified, suspended, or revoked. We request l
yet.r response within 60 days of receipt of this letter. This information is i
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required pursuant to 10 CFR 50.54(f) to assess conformance of PWks with their j
licensing basis and to determine whether additional hRC action is necessary.
l Our review uf information you submit is not subject to fees under the provision of 10 CFR 170.
It you choose te provide a portion of your response in association with your owners group, such action is acceptable.
This request for information was approved by the Office of Management and i;udget under clearance number 3150-0011 which expires December 31, 1989.
Lormients on burden and duplication may be directed to the Office of Management and Budget, Reports Managenient Roorr, 3?08, New Executive Of fice Building, Washingten C.C. 20503.
Sincerely, Frank J. Miraglia Associate Director for Projects Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
Enclosures:
As stated
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ENCLOSUPE : TO ATTACHMENT 1
- t.f0RMATIOfi PERTINFNT 10 LOSS OF RESIDUAL HEAT REMOVAL SYSTfMS WH!! r THE RCS 'S PARTIALLY FILLED hany maintenance and test activities Conducted during an rutap require lowering the water lesel in the reactor coolant system (RCS) to below the top j
of the reactnr vessel (RVI cr (as is dtre many timesl to the centerline elevation of the RV nozzles. This operating regime is sometimes known as "mid-loop" operation, it places unusual demands cn plant equipment and operntors because of narrow control margins and limitations associated with equipment, instrurnentation, procedures, training, and the ability to isolate containment. Difficulty in controlling the plant while in this condition often leads to loss of the residual heat removal (RHR1 system (Table 1).
Although this issue has been the topic of many cormiunications and investigations, such events contir.ue to occur at a rate of several per year.
Pecent knowledge has provided additional insight into these events. Although the full implicat %ns of this knowledge remain to be realized, our preliminary assessments have cleerly established real and potential inadequacies ossociated with operation while the PCS is partially filled. These include:
nut understanding the nuclear steam supply system (NSSS) response to loss of thP, inadequatt instrumentation, lack of analyses addressing the issue, lack of applicable procedures and training, and failure to adeouately address the safety impact of loss of decay heat removal capability.
The folicwing items are applicable to these conclusions:
(1) Plants enter an unanalyzed condition if boiling occurs following loss of RHR.
for example:
(a) Unexpected PCS pressurization car occur.
No pressurization would cccur with a water /steatr-filled RCS with water on the steam generator (SG) secondary side, because RCS steam
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would. ondense in the SG tubes and the condensate would return to the 14. Air in the PCS can block the flow of stearn through passages, such as the entrance portion of SG tubes, so that steam cannot reach cool surfaces.
Failure to cond ose the steam causes pressurization in the kCS until the air compresses enough for steam to reach cooled tube surfaces.
This pressurization occurred during the April 10, 19E7 event at Diablo Canyon since the PCS contained air.
Pressure-reached 7 to 10 psig, and would have continued to increase if RllR had not been restored. The operators began to terminate the event by allowing water to flow from the refueling water storage tank (RWST) into the RCS.
Increasing pressure would have elirninated 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 forcec cut of the RV, thereby reducing the tir,e between loss of RHP and initiation cf core damage.
This is a potential ccncern whenever there is an opening in the cold leg, such as may crist for repair of reactor coolant pumps (RCPs) or loop isolation valves.
Upper vessel / hot-leg pressurization could force the PV water level down with the displaced water lost through the cold-leg opening. A corresponding decrease in level would occur in the SG 51ce of the crossover pipes between the SGs and the RCPs.
This occurrence could be particularly serious if the cold-leg opening viere large or if makeup water flow to the PCS were small, as froni a chargir g purrp.
Cold-leg injection with elevated pressure in the upper vessel may not provide water to the core.
(N K5 water IEvel instrumentation nay provide ineccurate iriformation.
There are rtary facets to this issue.
Instrumentation may be indicating
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3 a level that differs from level at the RHR suction line, a temporary instrument may be in use that has no indication or alarms in the control rooni, and design and instal 4 tion deficiencies may ex.ist. We have observed the following:
(a) Connections to the RCS actually provide a water level indicatic' ap-stream of the RCP location. This water level is higher than the wcter level at the RHR suction connection because of flow fron the injection to the suction locations and because of entering water mcnientum, which increasee level on the RCP side of the cold-leg i
injection location, Ingestion of air at the RHP suction connection will result in transporting air into the cold legs; this can potentially increase l
l pressure in the air space in the cold legs relative to the hot legs, f
Level instrumentation may respond to such a pressure change as I
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 cccurs).
(b) l'se of long ler.gths of small-diameter tubing which can lengthen instrument response time ar.d cause. perturbations such as RCS pressure changes to appear as level changes, installation with tubing elevation changes which can trap air bubbles or water droplets, and installation which rakes it possible for tubing to be 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 because they share common connections.
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1 (d)' Tygon tube installations faintly marked at 1-foot intervals that -
have no prevision for holding the tube in place.
(e) Instrunntation in which critical inspections were nn+ perfonned after the installation.
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Instrurrentation in which no provisions were made to ensure a single phase in connection tubing or that tubing was not plugged.
1 (g) Use of instrumentation without performing an evaluation of indicated FCS 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 PCS level to plant elevatior 106' 6" innediately af ter steam generator tubes drained, and indications of erratic RHP pump current were observed.
Restoring the PCS level to 106' 10" was reported to have eliminated the problem. WHR operation was terminated a few hours later at an indicated level cf 107' 4" because the operators observed erratic kHR pump current indications. The licensee later reported that vortexing initiated under those conditions at 107' 5-1/2", and was fully developed at 107' 3-1/2".
Procedures in place at the time ui the event indicated the minimum allowable level to be 107' 0" (the hot-and cold-leg centerline elevation) or 107' 3".
(b) Additional phenornena appear to occur under bir ingestion conditions.
These include:
5 RHR pumps at Diablo Canyon were report 00 to handle several percer.t air with ro discernible flow or pump current change from that of singl(-phase operation.
A postulate is that cir in the RHR/ reactor ccolant system can migrate or redistribute, and thus cause level changes which are a+ voriance with those one would. expect.
This is a possible explanation for observed behavior in which lowering the HCS 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 RHR system being replaced by RCS Water.
Failure to understand such behavior leads operators to mistrust level iristrumentation and to perform operational errors.
(c) Operators typicelly will start another RHR pump if the operating pump is lost.
Experience and an understanding of the phenomena clearly shew triat less of the second pump should be expected. The cause of loss of the first pump should be well understood and nornally should be corrected before attempting to run another RHR pump.
(d) Typical operation wl.ile the RCS is partially filled provides a high EHR flow rate, which n,ay be required by TS, but which may be unnecessary under the unique conditions associated with the cartielly filled RCS. Air irgestion problems are less at low flow rates.
(4) Only limited instrumentation may be available to the aperator while the RCb is partially filled.
i 6-4 (a) l.evel 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) Typicali), PHR system temperature indication is the only temperature provided to the operators.
Loss of RHR leaves the operator with no hCS temperature indication.
This car result in a TS violation, as occurred at Diablo Canyon on April 10 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 bciling had initiated.
(c) RHR pump motor current and flow rate may not be alarmed and scales may not be suitable f or operation with a partially filled PCS.
(c) kHR suction and discFarge pressures may not be alarmed and scales may not be suitable for cperation with a partially filled RCS.
(5)
Licensees typically conduct operations while the RCS is partially filled, the containment equipment hatch has been renoved, and operations are in progress which impact the ability to isolate containment.
- Planning, procedures, and training do not address containment closure in response to ioss of RHP cr core damage events.
This is inconsistent with the
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sensitivity at sociated with partially filled RCS operation and the histor,v et loss of PHP under this operating condition.
i f> ) Licensees typically conduct test and maintenance operations that can perturb the PfS and PUE systen, while in a part ially filled FCS ccndition.
The sensitivity of the operation and the historical record i
indicete this is not prudent.
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L Table 1 37 LOSS-OF-DHR* EVENTS ATTRIBUTED TO INADEQUATE RCS LEVEL Docket Plant Date Duration Heatup 344 Trojan 05/21/77 55 min.
Unknown l
03/25/78 10 min.
Unknown I
10 min.
Unknown 04/17/78 Unknown Unknown i
334 Beaver valley 1 09/04/78 60 min.
145-175 F 366 Millstone 2 03/04/79 Unknown 150 208 F 272 Salem 1 06/30/79 34 min.
Unknown 334 Beaver Valley 1 01/17/80 Unknown Unknown 04/08/80 35 min.
None 04/11/80 70 min.
101-108 F 03/05/91 54 min.
102-168 F 344 Trojan 06/26/81 75 min.
140-150 F 369 McGuire 1 03/02/82 50 min.
105-130 F 339 North Anna 2 07/30/82 46 min.
Unknown 338 North Anna 1 10/19/82 36 min.
Unknown 10/20/82 33 min.
Unknown
%9 McGuire 1 04/05/83 Unknown Unknown 339 North Anna 2 05/03/83 Unknown Unknown 05/20/82 8 min.
Unknown 26 min.
Unknown 60 min.
Unknown 28U Surry 1 05/17/83 Unknown Unknown 328 Sequoyah 2 08/06/83 77 min.
103-195 F 370 McGuire 2 12/31/83 43 min.
Unknown 01/09/84 62 min.
Unknown 344 Trojan 05/04/84 40 min.
105-201 F 316 OC Cook 2 05/21/84 25 min.
Unknown 368 ANO-2 08/29/84 35 min.
140-205 F 295 Zion 1 09/14/84 45 min.
110-147'F 339 North Anna 2 10/16/84 120 min.
Unknown 413 Catawba 1 04/22/85 81 min.
140-175 F
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327 Sequoyah 1 10/09/85 43 min.
<1 F 296 lion 2 12/14/85 75 min.
s15 361 San Onofre 2 03/26/86 49 min.
114-210 F 36?
Waterford 3 07/14/86 221 min.
138-175 F 12/
Sequo.yan 1 01/28/87 90 min.
95 115 f l
323 Diablo Canyon 2 04/10/87 85 min.
100-220 F l
ENCLOSURE 2 TO ATTACllMENT 1 S S ! b'S No..: 6835 l
1N 87-23
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l UNITED STATES l
NUCLEAR REGllLATORY COPylSS10N l
OFFICE OF NUCLEAR REACTOR REGULATION WASHlHGTON,' O.C.
20555 Vay 't7,1987 5
.l NPC. INF09 PATI 0h N0llCE NO. F7-?3: LOSS OF DECAY HEAT REMOVAL DURING LOW REACTOR C00LAMT LEVEL OPEPAT!ON l
I Addressees.
All holders or an operating license or a construction permit for pressurized-I water reactor facilities.
Purpose:
This notice provides information reoarding the loss of decay heat removal capability at pressurized water reactors resulting from the loss of PFP pump l
suction during plant operations with low reactor coolant levels. it is ex-petted that recipients will review this information for applicability to their reactor facilities arid 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.
Nscriptinn of Circur. stances:
On April 10, 1987 the Diablo Canyon Unit ? reactor experienced a loss of decay heat removal capability in both trains.
The reactor coolant system had been i
drained down to the mid-height of the hot-leg piping in preparation for the rer.caal 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 opering in the head, water was spilled from the partia'ly unsealed manways, and the airborne radioactivity levels in the containment rose above the maximum perTnissible concentration of noble gases allowed by 10 CFR 00.
The reactor, which was undergoing its first refueling, had been s hut down for seven days at the time and the containment equipment ba tch had beer, opened.
Erroreous level instrumentation, inadequate knowledge of pump suction head / flow requirements, incomplete assessment of the behavior of the air / water mixture in the systen and poor coordination between control room operations and contain-ment activities all contributed to the event.
Under the conditions that eristed, the system that indicated the level of coolant in the reactor vessel re a d hi gh" and responded poorly 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 comrnence. 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.
P.705?00749
IN 87-?3 Pay ?7, 1987 Pace 2 of 5 TFe event Ngan 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 isolated. Hnwever, because of a leaking boundarv valve, this actipn 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 but 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 electric.al current of the active RHR pump (No, ? ?) was observed to be fluctuating. The ?-l pump was started j
and the 2-2 pump was shut down.
However, the current on the ?-l pump also i
fluctuated, 50 it was imediately 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 leakage, the true vessel level, or the status of the work. on the steam generator manways. Operators were sent to vent the RHP purps.
One pump was reported to be vented at 10:0?
p.m.
At 10:21 p.m. an attempt was made to start this RHP pump, but the current 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 ar.c personnel began to evacuate from the containment buildine.
At 10:3F. p.m. when the operators learned that the steam generator manways had r.ct been recoved, 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 PHP purps. The indicated RHR pump discharoe temperature imediately rese to
??0* F.
At this tir'e 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 documented 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 Infomatior Notice 86-101 describes four such events that occurred in 1985 ar.d 1986. NRC Case Study Report AE0D/C503 describes six such events that occurred in 1984, five that occurred in 1983, and seven that occurred in 1982.
IE Information Notice F1-09 described an event at Beaver Valley in March 1981. The case study report further indicates that a total of 32 such events occurred from 1976 through 1004 The documentation includes descriptions of a total of 23 events that have occurred since 1981 involving loss of decay beat removal capability resultine from a loss of pump suction while operating at reduced water levels.
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IN 87-23 l
Fay 77, 1987 l
Page 3 of 5 For al' but four of these 23 events the primary cause of the loss of pump 1
i suctior and loss of decay heat removal capability was attributed to incorrect, ir acc: eate, or iradequate level indication.
Two events were attributed to loss I
of purp suction because of vortexing brought on by the simultaneous operation of both pumps.
In many of these events procedural errors were also a contrib-oting 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 problen, switched to the second pump. There are repeated references to dif ficulties in getting the pumps vented quickly af ter air binding had nccurred and to the operators' inability to take imediate action to raise reactor vessel levels until the safety of personnel working or the primary systems could be assured. The length of time that decay heat removal was completely lost varied frne eight minutes to two hours and averaged almost an hour.
In at least three previous cases, boiling is known to have occurred.
I A number of actions have been recomended previously to prevent the loss of PHP l
j pump suction during low vessel level operations.
These include' l
Providing accurate level instrumentation designed for reduced vessel water level operations.
Providing alarms in the control room for low decay heat removal flow and low water level.
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including in the procedures specific require: ants for frequent monitoring and strict limits on level.
Considering in the procedures the possibility of vortex formation and air entrainment, including a precaution against starting a second RHP pump until the cause of the loss of the first pump is determined and corrective actions have been taken.
Training the operators on the correlation between water level and pump l
speed at the onset of vortexing and air entrainment.
Careful planning, coordination, and communication with control room persenr.el regarding all ongoing activities which could affect the primary system inventory.
The NDC review of the Diablo Canyon event indicated that vortexing ano air entrairmert may occur at higher water levels than anticipated.
In addition, operat' ion at mid-hot-leg levels can lead to unanticipated conditions which may not have been adequately considered in instrumentation design and procedure preparation.
The NPC staff's initial assessmer.t of this event has identified the potential for a 5ignificant loss of decay heat removal capability both from a total loss of the PHR system and from a loss of the steen cenerator heat sink due to air blanketino of the steam ger.erator tubes. Correct operator actions then become criticel fnr plant recovery.
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m, IM 87-?3 May 27, 1987 Page 4 of 5 NPC co runic.'tions in the past have expressed serious concern with failures to maintain adeouate decay heat removel capability.
!E Inforretion Notice 81-09 pointed out that loss of shutdown cooling capatility had been found to be a potentially significant contributor to the total risk.
AE00/C503 (nd other sources irdicate that the time available to restore shutdown cooling before core uncovery can occur is not necessarily large. At four days af ter shutdown from long-tere power operation, with the vessel drained down to the' RHP suction loss level, the vessel water can heat to the boiling point in about 1/2 hour.
j Under such conditions boiloff to the core uncovery level can occur in less than j
two hours.
following the loss of decay heat removal capability on April 10, 1987 at Diablo Canyon, PG!E took a number of actions to prevent loss of RHP suction during low level operation and to improve recovery should such a loss occur.
These actions included the following:
Esaluation of the reactor vessel level indicatino system to determine the level at which vortexing would occur and the effect of vortexing on the level measurement.
Enhancements of the instrumentation to include accurate level measurement, alarr capability and core exit temperature measurement during low level operation.
Enhancement of procedures to include requirements 'cr verifying proper RHR pump suction before starting the second RHR pump. Also included are precautions specifying minimum vessel levels as a function of PHR flow.
Improvements in worf planning, control and communication to include a restriction of the work scope to items that do not have the potential to reduce RCS inventory.
Improvement of operator traird ng including a discussion of the potential causes of RHR flow Icss, as well as recovery procedures.
The hPC is cur ently considering additional generic action on this issue,
l 1H 87-23 May 27, 1987 Page 5 of 5 This infomation notice requires no specific action or written response.
If ycu have any cuestions about this matter, please contact the Regional Adr.inistrator o' the appropriate recional office or this office.
1 fdw w
Charles E Rossi, Director Division of Operational Events Assessr.ent
, Office of Nuclear Reactor Regulation Techrical Contacts: Donald C rirkpatrick, NPR (301) 492-8166 i
Warren C. 1. yon, NRR (301) 4c?-7605
Attachment:
1.
List of Recently issued NPC Infomation Notices l
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l N
+ 9,tAcnment I h 87-23 May
'7, 1087 3
LIST OF DECENTLY ISSUED lhf0PMAT10N NOTICE". 1Q87 Inforn tion Date of Notice No.
Subiect issuance issued to 87-??
Operator Licensing Dequali-5/22/87 All research and fication Examinations at nonpower reactor Nonpower Peactors facilities.
87-?!
Shutdown Order issued Recause 5/11/87 All nuclear power Licensed Operators Asleep facilities holding While on Duty an OL or CP and all licensed operators.
i F7 'T Hydrogen leak in Auxiliary 4/20/87 All nuclear power Building facilities holding an OL or CP F6-100 Degradation of Reactor 4/20/87 All PWD facilities Sup. 1 Coolant System Pressure holding an OL or CP.
Boundary Resulting from Boric Acid Corrosion 96-6.
Deficiencies in Upgrade 4/20/87 All nuclear power Sup. 1 Programs for Plant facilities holding Emergency Operating a CP or OL, Procedures.
85-61 Hisadministrations to 4/15/87 All licensees Sup. 1 Patients Undergoing Thyroid authorized to use Scans byproduct material 87-19 Perf oration and Cracking of 4/9/87 All Westinghouse Rod Cluster Control Assemblies power PWR facilities j
]
87-18 Unauthorized Service on 4/8/87 All NRC licensees j
Teletherapy Units by Non-authorized to use licensed Maintenance Personnel radioactive material in teletherapy Units 87-17 Response Time of Scram a/7/87 All GE BWP facilities Instrument Volume Level holding an OL or CP Detectors OL = Operating License CP = Construction Pennit
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