NRC Generic Letter 1987-12

NRC Generic Letter 1987-012: Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) Is Partially Filled

ML031150504
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Washington Public Power Supply System, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Clinch River, Crane
Issue date:	07/09/1987
From:	Miraglia F J Office of Nuclear Reactor Regulation
To:
References
GL-87-012, NUDOCS 8707100112
Download: ML031150504 (12)

e ---i$J .1 4 #UNITED STATES NUCLEAR REGULATORY

COMMISSION

WASHINGTON.

D. C. 205 July 9, 1987 ALL LICENSEES

OF OPERATING

PWRS AND HOLDERS OF CONSTRUCTION'

PERMITS FOR PWRS Gentlemen:

SUBJECT: LOSS OF RESIDUAL HEAT REMOVAL (RHR) WHILE THE REACTOR COOLANT SYSTEM (RCS) IS PARTIALLY

FILLED (GENERIC LETTER 87-12 )Pursuant to 10 CFR 50.54(f), the NRC is requesting'information to assess safe ,operation of pressurized-water reactors (PWRs) when the reactor coolant system (RCS) water levelis below the top of the reactor vessel (RY). The principal concerns are (1) 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; and (3)whether any threat to safety that warrants further NRC attention exists in this condition.

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 facilities toensure 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 damage, and could have withstood the loss-of-RHR

condition for more than a day with no operator action, slightly different conditions could have led to an accident involving core damage 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 repair to a reactor coolant pump (RCP) or to a loop isolation valve. The lost water would no longer be available to cool the core, and if makeup water were 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 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 removed, and by the lack of procedures that address prompt containment isolation 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 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 licensing basis. Your description is to include the following:

1OO 112 m> toSoC a

2 (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 with the RCS partially filled, including any interlocks that could cause a disturbance to the system. Examples of the type of information required are the time between full-power operation and reaching a partially filled condition (used to determine decay heat loads); requirements for minimum steam generator (SG) levels; changes in the status of equipment for maintenance and testing and coordination of such operations while the RCS is partially filled; restrictions regarding testing, operations, and maintenance that could perturb the nuclear steam supply system (NSSS); ability of the RCS to withstand pressurization if the reactor vessel head and steam generator manway are in place;requirements pertaining to isolation of containment;

the time required to replace the equipment hatch should replacement be necessary;

and requirements pertinent to reestablishing the integrity of the RCS pressure boundary.(2) A detailed description of the instrumentation and alarms provided to the operators for controlling thermal and hydraulic aspects of the NSSS during operation with the RCS partially filled. You should describe temporary connections, 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 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 after the RHR function may be lost.(3) Identification of all pumps that can be used to control NSSS inventory.

Include: (a) pumps you require be operable or capable of operation (include information about'such pumps that may be temporarily removed from service for testing or maintenance); (b) other pumps not included in item a (above); and (c) an evaluation of items a and b (above) with respect to applicable TS requirements.

(4) A description 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.

(5) 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 draindown 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, treatment of boiling in the core with and without RCS pressure boundary integrity, calculations of approximate time 0

VI- 3 -from loss of RHR to core damage, level differences in the RCS and the effect upon instrumentation indications, treatment of air in the RCS/RHR system, Including the impact of air upon NSSS and instrumentation response, and treatment of vortexing at the connection of the RHR suction 1ine(s) to the RCS.Explain.how your analytic 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 A operations that may perturb the NSSS, including restrictions upon testing,.maintenance, and coordination of operations that could upset the condition of the NSSS; and (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 operation while the RCS is partially filled. 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 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.

(8) Comparison of the requirements implemented while the RCS is partially filled and requirements used in other Mode 5 operations.

Some requirements and procedures followed 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 scheduled to be made.Enclosure

1 contains insight which experience indicates should be well understood before commencing operation with a partially filled RCS. Your response to this 50.54(f) letter request should encompass the topics contained in Enclosure

1. Additional information is contained in the NRC Augmented Inspection Team report, NUREG-1269, "Loss of Residual Heat Removal System, Diablo Canyon Unit 2, April 10, 1987." A copy of NUREG-1269 is enclosed.

-4-Your response addressing items 1 through 9 (above) is to be signed under oath or affirmation, as specified in 10 CFR 50.54(f), and will be used to determine whether your license should be modified, suspended, or revoked. We request your response within 60 days of receipt of this letter. This information is required pursuant to 10 CFR 50.54(f) to assess conformance of PWRs with their licensing basis and to determine whether additional NRC action is necessary.

Our review of information you submit is not subject to fees under the provision of 10 CFR 170. If you choose to 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 Budget under clearance number 3150-0011 which expires December 31, 1989.Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management Room 3208, New Executive Office Building, Washington D.C. 20503.Sincerely, Frank J. Miraglia 1r£Associate Director 'or Projects u Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Enclosures:

As stated ENCLOSURE

1 INFORMATION

PERTINENT

TO LOSS OF RESIDUAL HEAT REMOVAL SYSTEMS WHILE THE RCS IS PARTIALLY

FILLED Many maintenance and test activities conducted during an outage require lowering the water level in the reactor coolant system (RCS) to below the top of the reactor vessel. (RV) or (as is done many times) to the centerline elevation of the RV nozzles. This operating regime is sometimes known as"mid-loop" operation.

It.places unusual demands on plant equipment and operators because of narrow control margins and limitations associated with equipment, instrumentation, 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 (RHR) system (Table 1).Although this issue has been the topic of many communications and investigations, such 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 operation while the RCS is partially filled. These include: not understanding the nuclear steam supply system (NSSS) response to loss of RHR, inadequate instrumentation, lack of analyses addressing 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 example: (a) Unexpected RCS pressurization can occur.No pressurization would occur with a water/steam-filled RCS with water on the steam generator (SG) secondary side, because RCS steam

-2 -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 pressurization in the RCS until the air compresses enough for steam to reach cooled tube surfaces.

This pressurization occurred during the April 10, 1987 event at Diablo Canyon since the RCS contained air. Pressure reached 7 to 10 psig, and would have continued to increase if RHR 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 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 col'd leg, such as may exist for repair of reactor coolant pumps (RCPs) or loop isolation valves. Upper vessel/hot-leg pressurization could force the RV water level down with the displaced water lost through the cold-leg opening. A corresponding decrease in levet would occur in the SG side of the crossover pipes between the SGs and the RCPs.This occurrence could be particularly serious if the cold-leg opening were large or if makeup water flow to the RCS were small, as from a charging pump. Cold-leg injection with'elevated pressure in the upper vessel may not provide water to the core.(2) RCS water level instrumentation may provide inaccurate information.

There are many facets to this issue. Instrumentation may be indicating r ?I1-3 -a level that differs from level at the RHR suction instrument may be in use that has no indication or room,.and design and installation deficiencies may observed the following:

line, a temporary alarms in the control exist. We have (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 because of flow from the injection to the suction locations and because of entering water momentum, 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; this 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 long lengths of small-diameter tubing which can lengthen.instrument

• response time and 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 makes 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.

-4-(d) Tygon tube installations faintly marked at 1-foot intervals that have no provision for holding the tube in place.(e) Instrumentation in which critical inspections were not performed after the installation.(f) Instrumentation in which no provisions were made to ensure 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" immediately after steam generator tubes drained, and indications of erratic RHR pump current were observed.

Restoring the RCS level to 106' 10" was reported to have eliminated the problem. RHR operation was terminated a few hours later at an indicated level of 107' 4" because the operators observed erratic RHR pump current indications.

The licensee later reported that vortexing initiated under those conditions at 107W 5-112", and was fully developed at 107' 3-1/22". Procedures in place at the time of the event indicated the minimum allowable level to be 107' 0" (the hot- and cold-leg centerline elevation)

or 107' 3".Cb) Additional phenomena appear to occur under air ingestion conditions.

These include:

-5-0 RHR pumps at Diablo Canyon were reported to handle several percent air with no discernible flow or pump current change from that of single-phase operation.

o A postulate 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 in which lowering the 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 RHR system being replaced by RCS water.Failure to understand such behavior leads operators to mistrust level instrumentation and to perform operational errors.(c) Operators typically will start another RHR pump if the operating pump'is lost. Experience and an understanding 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 should'be'corrected before attempting to run another RHR pump. --(d) Typical operation while the RCS is partially filled provides a high RHR flow rate, which may be required by TS, .but which maybe unnecessary under the unique conditions associated with the partially filled RCS. Air ingestion problems are less at low flow rates.(4) Only limited instrumentation may be available to the operator while the RCS is partially filled.

-6-(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 temperature indication is the only temperature provided to the operators.

Loss of RHR leaves the operator with no RCS temperature indication.

This can 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 boiling had initiated.(c) RHR pump motor current and flow rate may not be alarmed and scales may not be suitable for operation with a partially filled RCS.(d) RHR suction and discharge pressures may not be alarmed and scales may not be suitable for operation with a partially filled RCS.(5) Cicensees typically conduct operations while the RCS is partially filled, the containment equipment hatch has been removed, and operations are 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 partially filled RCS operation and the history of loss of RHR under this operating condition.

(6) Licensees typically conduct test and maintenance operations that can perturb the RCS and RHR system while in a partially filled RCS condition.

The sensitivity of the operation and the historical record indicate this is not prudent.

-r ---1 p F%, *: -; 7S w .Table 1 37 LOSS-OF-DHR*

EVENTS ATTRIBUTED

TO INADEQUATE

RCS LEVEL Docket 344 334 366 272 334 344 369 339 338 369 339 Plant Date Trojan Beaver Valle Millstone

2 Salem 1 -Beaver Valle.Trojan McGuire 1 North Anna 2 North Anna 1 f McGulre 1 North Anna 2 05/21/77 03/25/78 280 328 370 344 316 368 295 339 413 327 296 361 382 327 323 Surry 1 Sequoyah 2 McGuire 2 Trojan DC Cook 2 ANO-2 Zion 1 North Anna 2 Catawba 1 Sequoyah 1 Zion 2 San Onofre 2 Waterford

3 Sequoyah 1 Diablo Canyo 04/17/78 y 1 09/04/78 03/04/79 06/30/79 y 1 01/17/80 04/08/80 04/11/80 03/05/81 06/26/81 03/02/82 07/30/82 10/19/82 10/20/82 04/05/83 05/03/83 05/20/82 05/17/83 08/06/83 12/31/83.01/09/84 05/04/84 05/21/84 08/29/84 09/14/84 10/16/84 04/22/85 10/09/85 12/14/85 03/26/86 07/14/86 01/28/87 n 2 04/10/87 Duration 55 min.10 min.10 min.Unknown 60 min.Unknown:34 min.Unknown 35 mMn.70 min.54 min.75 min.-:50 min.46 min.36 mmn.-33min.Unknown Unknown.8 mn.-26 min.-t60 min.Unknown 77 min.43 min..62 min.; 40 min.25 min.35 mdb1 .,t 45 min. 7 120 min.81 min.43 min.75 min.49 min.221 min.90 min.85 min.Heatup Unknown Unknown Unknown Unknown 145-175°F 150-208°F Unknown Unknown None 101-108°F 102-168°F 140-150°F 105-130°F Unknown Unknown Unknown Unknown Unknown'Unknown Unknown Unknown Unknown 103-195°F Unknown Unknown 105-201°F-'Unkhown si-t~-e140`-2050°F-;3

11 110-17°F Unknown*140.175 0 F<1 0 F 114-210°F 138-175°F 95-115°F 100-220°F* Decay heat removal V--L .U 1 ;T LIST OF RECENTLY ISSUED GENERIC KsTERq to 4.-..--.I.L.ttm NO, subject Dats m4 Isasuance Issued To IL *7-;2 0.54 #1 LETTER RN. LOSS OF; U SDUAL HEAT REMOVAL (RHR DURINO MID-LOOP 0PERATION 3L 37-11 RELAxATION

IN ARBITRARY INTERMEDIATE

PIPE RUPTURE REOUIREMENTS

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