ML20056E936

Discusses Safety Significance of Corrective Actions Taken by Licensee in Response to IE Bulletin 86-003

ML20056E936
Person / Time
Site:	Point Beach
Issue date:	08/09/1993
From:	Rosenthal J NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:	Berkow H, Butler W, Hannon J Office of Nuclear Reactor Regulation
References
IEB-86-003, IEB-86-3, NUDOCS 9308250311
Download: ML20056E936 (3)

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MEMORANDUM FOR: John N. Hannon, Director  :

Directorate Ill-3 Division of Reactor Projects-III/IV/V  !

Office of Nuclear Reactor Regulation l l

Walter R. Butler, Director  !

Directorate I-3 Division of Reactor Projects-I/II  :

Office of Nuclear Reactor Regulation l

Herbert N. Berkow, Director  !

Directorate 11-2  :

Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation FROM: Jack E. Rosenthal, Chief Reactor Operations Analysis Branch  !

Division of Safety Programs Office for Analysis and Evaluation ,

of Operational Data  ;

i SUBJECT- SAFETY SIGNIFICANCE OF CORRECTIVE ACTIONS i TAKEN BY POINT BEACH IN RESPONSE TO j IE BULLETIN 86-03 i

After reviewing Licensee Event Report (LER)92-010 (Enclosure 1) for Point Beach l Nuclear Plant (PBNP) Units 1 and 2, AEOD is concerned with the reliance on operator i actions outside the control room to establish suction from the containment sump for the residual heat removal (RHR) pumps and the safety injection (SI) pumps after a loss-of-coolant accident (LOCA). The need to rely on operator action resulted from  ;

commitments made in response to IE Bulletin 86-03, " Potential Failure of Multiple ECCS Pumps Due To Single Failure of Air-Operated Valve In Minimum Flow  !

Recirculation Line" (Enclosure 2). Two other sites, Ginna and Turkey Point, appear to  ;

have made similar short-term commitments. Based on current individual plant i evaluation models, could a permanent modification be made that would eliminate the -i need to take operator actions outside the control room and thereby reduce the potential l for core damage? i To address the concern of SI pump damage due to valve closure in the minimum flow recirculation line to the refueling water storage tank (RWST), PBNP modified the valves i on the minimum flow line for the SI pumps from normally open (failing closed on loss of  ;

nonsafety-related air and loss of electrical power) to gagged open (Enclosure 3). This - I ensures the minimum flow line remains open so the pump does not become dead headed I l

and damaged after a LOCA. As discussed in the LER, the minimum flow valves are  !

interlocked with the containment sump suction valves and must close before the 8 1 j 388 988u 888"8366-P NRC FtLE CENTER COPY j PDR ,

Multiple Addressees  !

containment sump suction valves can be opened to prevent sump water flow to the RWST.

As discussed in LER 88-009 dated December 16,1988, (Enclosure 4), PBNP discovered  :

that insufficient time was allowed for operator action in Emergency Operating Procedure 13," Transfer to Containment Sump Recirculation." This procedure directed the operator to perform the final steps of establishing containment sump recirculation when ,

the RWST level decreased to 10 percent. At 6 percent RWST level, a caution in the procedure required that all pumps which are taking suction from the RWST be secured.  ;

As a result of an audit, calculations were performed and a determination made that following a large break LOCA, there was insufficient time for operator action based on the procedural steps. A commitment was made to revise the procedure.  !

LER 89-004 dated May 19,1989, (Enclosure 5) describes the discovery of a previous analysis inadequacy which impacts the time available for manual actions necessary to l effect the switch-over of the RHR and SI pumps from the injection phase to the ,

containment sump recirculation phase while one train is out of service and during a large break LOCA. Because of steam entrainment and voiding in the core, it was estimated  :

that core uncovery could occur in~ 1.5 to 3 minutes after the cessation of all RHR and SI flow. The LER states that the transfer time observed in walk-throughs and simulator runs is in the order of 2-4 minutes. It is not clear whether this includes the operator l actions that need to be taken outside the control room, including removing the gags on the minimum flow valves. Long term corrective actions were stated to include further  :

analysis and possible plant modification. l Based on LER 92-010 dated January 7,1993, it is unclear whether PBNP has revised the i procedure as indicated in LER 88-009, or completed the further analysis and possible  !

plant modifications indicated in LER 89-004. AEOD is not convinced that the responses j to IE Bulletin 86-03 (Enclosure 2) have considered all accident scenarios, particularly a large break LOCA, with respect to long term corrective action. Since the development  :

of the plant specific individual plant evaluation, has the potential for core damage following a large-break LOCA been evaluated based on the probability of the operators i failing to close the minimum flow line isolation valves, and thereby failing to'open the ,

containment sump isolation valves to supply flow in sufficient time?

Orkinal Signed by: h Sem -

t Jack E. Rosenthal, Chief  !

Reactor Operations Analysis Branch Division of Safety Programs ,

Office for Analysis and Evaluation i of Operational Data j

Enclosures:

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Document Control Desk U.S. NUCLEAR REGULATORY COMMISSION Mail Station P1-137 Washington, DC 20555 Gentlemen:

DOCKETS 50-266 AND 50-301 LICENSEE EVENT REPORT 92-010-00 ISOLATION OF SI PUMP FLOW PATH DURING INSERVICE TESTING OF MINIMUM FLOW RECIRCULATION LINE ISOLATION VALVES 1 POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 Enclosed is Licensee Event Report 92-010-00 for Point Beach Nuclear Plant, Units 1 and 2. This report is provided in accordance with 10 CFR 50.73 (a) (2) (v) (D) , "The licensee shall report...any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to miti-gate the consequences of an accident."

This report describes the possible isolation of all available flow paths for the safety injection pumps during performance of Inservice Tests IT-40, " Safety Injection Valves (Quarterly),

Unit 1," and IT-45, " Safety Injection Valves (Quarterly), Unit 2."

The isolation of all available flow paths could result in operating the pumps at shutoff head, ultimately damaging the pumps and rendering them inoperable.

Please contact us if any further information is required.

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I AP S* PACT At 1450 on Decer.ber 8,1992, while Point Beach Nuclear Plant (PBNP) Units 1 and 2 were '

4 operating at 100% and 95% power respec.ively, it was discovered that Inservice Tests IT-40, " Safety Injection valves (Oaarterly), Unit 1,* and IT-45, " Safety Injection valves (Quarterly), Unit 2,* could lead to the isolation of all available flow paths for the ,

, safoty injection (SI) pumps. Tests IT-40 and IT-45 perform quarterly otroke tests of l oafety in$eetion/ containment spray minimum flow recirculation line isolation valves 1&251-897A and 1&2SI-897B (hereinaf ter referred to as valves 897A&B). IT-40 and IT-45 place the plant in a condition in which pump damage could occur if the SI pumps automatically started while reactor coolant system (RCS) pressure was greater than SI purep shutoff head and either Valve 897A or valve 8973 remained shut. Operating the SI pumps at shutof f head would cause pump damage af ter approximately one minute. The tests were last perforced on 11/15/92 (Unit 1) and 11/19/92 (Unit 2). A 4-hour NRC ENS notification was made in accordance with 10 CFR 50.72(b)(2)(iii)(D). The NRC Resident Inspector was also notified. A Probabilistic Risk Assessment (PRA) was subsequently performed and deter =ined that the probability of this event occurring is approximately .

1.0 E-6 events / year, or an increased pu=p damage risk of approximately 2 percent. Due

to the increased risk of damaging the SI pumps by testing Valves 897AEB on a quarterly  ;

frequency, the tests will subsequently be perfore,ed on a cold shutdown frequency. ,

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01 1 l 0 0IO O I2 O I4 fort a ea w. w ==c s.- asuw Tn EVENT DTSCRIPTION At 1450 on December 8, 1992, while Point Beach Nuclear Plant (PBNP) Units 1 and 2 were operating at 100% and 95% power respectively, it was discovered that Inservice Tests IT-40, " Safety Injection valves (Quarterly), Unit 1," and 1T-45, " Safety Injection Valves (Quarterly), Unit 2,* place the plant in a condition in which damage could occur to both SI pumps. IT-40 and IT-45 perform quarterly stroke tests of SI/CS mini-recirculation lina isolation valves 1&2SI-897A and 1&2SI-897B (hereinafter referred to as Valves 897A&B). The damage could occur if the SI pumps automatically started while either Valve 897A or 897B was shut. Operating the SI pumps with either Valve 897A or 897B shut would cause SI pump damage due to operation of the SI pumps at shutoff head without minimum recirculation flow if reactor coolant system (RCS) pressure was greater than SI pump shutoff head and plant operators failed to open one of the valves, 897A or 897B, within approximately one minute. The tests were last performed on 11/15/92 (Unit 1) and 11/19/92 (Unit 2). Upon identification of this condition on December 8, 1992, a 4-hour NRC ENS notification was made in accordance with 10 CFR 50.72(b)(2)(iii)(D). The NRC Resident Inspector was also notified.

Although the ens notification identified that the containment spray (CS) pumps could aloo be damaged under the same circumstances, this condition is now considered to be of less concern. The CS pumps have a flow path to containment regardless of the position of Valves 897A&B unless both of the two parallel motor-operated discharge valves per CS pump fail to open on an automatic signal. Because the CS pump discharge valves are powered from separate safeguards trains, concurrent failure of both pairs of discharge valves is not a credible event.

A Probabilistic Risk Assessment (PRA) was subsequently performed and determined that the probability of an automatic initiation of SI occurring while either Valve 897A or 897B is chut is approximately 1.0 E-6 events / year, or an increased pump damage risk of approximately 2 percent.Section II of the ASME Boiler and Pressure Vessel Code, Article IW-3412a, 1986 Edition, allows plants to identify those valves which cannot be tested during plant operation and provide for full-stroke testing of these specific valves during cold shutdowns. The PBNP Inservice Testing (IST) program accounts for valves requiring this type of testing in Appendix G, " Cold Shutdown Justifications.*

Th3refore, due to the elevated risk of pump damage while testing Valves 897A&B during plant operation, testing of valves 897A&B will be deferred to periods when the rospective unit is in cold shutdown and both SI pumps may be taken out of service.

Wioconsin Electric addressed a related issue in a letter to the NRC dated July 24, 1985.

The letter, submitted in accordance with 10 CFR 21, notified the NRC that the f ailure of a oingle eceponent in the control circuitry for the SI recirculation path isolation valves could, under specific circumstances, result in the failure of both safety injoetion pumps. During a post-implementation. review of the Emergency operating Procedures (EOPs),'it was discovered that a failure of the power supply breaker in the remote control circuitry.for Valves.897A&B would result in those valves closing. This f ailure would simultaneously' result in loss of valve position indication and defeat the tnnunciation for 897A&B valve closure on the main control board.- The corrective actions specified in response to this issue included gagging the manual handwheel operators on Valves 897ALB in the open position so that the automatic operators would be overridden.

This corrective action was also referenced in our response to NRC IEB 86-03,

• Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air-operated valve in the Minimum Flow Recirculation Line," dated November 12, 1986. However, quarterly inservi;e otroke tests in which the valves were ungagged for a short period of time and rapositioned for testing were considered at that time to pose no significant increase in rick to the SI pumps. -

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EQUIPMENT DESCRIPTION (Note: Information in [] indicates Energy Industry Identification l System (EIIS) identifiers)

An orificed minimum flow bypass line is provided at the discharge of each SI (BQ) pump l

[P) to recirculate flow to the refueling water storage tank (RWST)[TK) through a coenon  !

headar (or, " mini-recirc" line) in the event the pumps are run while the RCS [AB) l pressure is above the pumps' shutoff head. These bypass lines also permit the performance of periodic surveillance tests required by the Technical Specifications to prove pump operability. The recirculation line is provided with air-operated isolation Valves 897A&B [ISV), in series, which are closed to prevent flow of contaminated water ,

to the RWST when in the containment sump recirculation phase following an accident. .

Bacause Valves 897ACB f ail shut, theyrare snormally gagged open to prevent-closure on a loss of instrument air. If the SI pumps are operated without a flow path, the pumps will overheat and quickly deteriorate.

Valves 897AGB are interlocked with containment sump "B" isolation valves 1&2SI-851AGB

[ISV) (hereinafter referred to as Valves 851A&B). These motor-operated gate valves are normally closed except when required for containment sump recirculation following an accident. This interlock insures that Valves 851A&B cannot be opened until at least one velve, 897A or 897B, is closed which prevents the inadvertent release of contai.nment oump vapor or liquid to the RWST during the containment sump recirculation phase of long-term cooling following a design basis accident.

The manual handwheel operators on Valves 897A&B are currently maintained in the open position to prevent closure on a loss of instrument air.

CAUSE Th2 re-evaluation of the PBNP quarterly inservice testing practices for Valves 897AEB was prompted by INPO Nuclear Network Message OE 5692, " Loss of All ECCS Pumps During Monthly Surveillance Testing," transmitted on November 24, 1992, by Calvert Cliffs Nuclear Plant.

Prior to this re-evaluation, quarterly inservice stroke tests in which Valves 897AGB were ungagged for a short period of time and repositioned for testing were deemed nseassary and considered to pose no significant increase in risk to the SI pumps.

CORRECTIVE ACTIONS A. Immediate:

1. Purther testing of Valves 897AGB was suspended. l B. Short term:

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1. A Probabilistic Risk Assessment (FRA) was performed to determine the i probability of an SI actuation during the time Valves 897AGB are being tested. Station logs were reviewed and indicated that the approximate time to complete IT-40/45 is on the order of two hours (however, the valves are not shut for the full duration of the test) and therefore the time that the valves are ungagged each calendar quarter is small. Given this infornation, ,

it was determined that the probability of this event occurring is l approximately 1.0 E-6 events / year, or an increased pump damage risk of  !

approximately 2 percent. ,.

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2. The pumps' manufacturer, Byron Jackson, was consulted and stated that the SI  ;

pwtps can be operated at shutoff head for up to one minute before pump ,

degradation begins. Therefore, control operators would have up to one minute i after an automatic pump start to restore the flow path-if instrument air is -

available. If instrument air is not available, the valves would require manual handwheel operation. '

C. Long Tent:

1. A Cold Shutdown Justification (CSJ) for Valves 1&2SI-897A&B will be included ,

in the IST program to allow testing on a cold shutdown frequency. This change was submitted to the NRC on December 23, 1992.

2. Test procedures will be developed to provide for the inservice testing (stroke time, fail-safe, position indication verification, leak rate testing) of Valves 1&2SI-897A&B on a cold shutdown frequency. This will be completed by the operations Group by February 28, 1993.

3. The operations Group will ensure that all other valves currently tested under Procedures IT-40 and 17-45 on a quarterly basis will continue to be tested on a quarterly basis. Procedures to accomplish this testing will be developed if necessary. All necessary procedure revisions will be coepleted or new procedures developed by Tebruary 28, 1993.

U POPTABILITY This event is being reported under the requirements of 10 CTR 50.73(a)(2) (v)(D), "The licensee shall report...any event or condition that alone could have prevented the '

fulfillrrent of the safety function of structures or systems that are needed to mitigate 4

the consequences of an accident." A 4-hour NRC ENS notification was made in accordance 4

with 10CTR50.72 (b) (2 ) (iii) (D) . The NRC Resident Inspector was also notified.

SArrTY ASSESSMENT A Probabilistic Risk Assessment (PRA) was performed and determined that the probability of op* rating an SI pump at shutof f head af ter Valves 897AEB have been ungagged and the valves have failec shut is 1.0 E-6 events / year. The probability of failure of the SI pumps for other reasons is calculated to be 5.3 E-5 events / year. Therefore, this identified condition would result in an increased risk of failure of the SI pumps of cbout 24. This condition is a small contributor to the failure of the SI pumps. Hence, the probability of pump damage occurring as a result of the scenario described above is  ;

determined to not be a significant contributor to core damage frequency. The safety of  ;

the plant and the health and safety of the public and plant employees were not joopardized by this plant condition.

l CENEFIC IWPl.ICATIONS No generic irplications have been identified.

SIMILAR OCCURRENCES ,

Th2re have been no similar occurrences identified at PENP. _-

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SSINS No.: 6820 OMB No.: 3150-0011 IEB 86-03 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMEN' WASHINGTON, DC 20555 October 8, 1986 IE COMPLIANCE BULLETIN NO. 86-03: POTENTIAL FAILURE OF MULTIPLE ECCS PUMPS DUE TO SINGLE FAILURE OF AIR-OPERATED VALVE IN MINIMUM FLOW RECIRCULATION LINE Addressees:

All facilities holding an operating license or a construction permit.

Purpose:

The purposes of this bulletin are (1) to inforTn addressees of single failures of minimum flow recirculation lines containing air-operated isolation valves which could result in a corrnon-cause failure of all emergency core cooling system (ECCS) pumps in a system, (2) to request that licensees affected by the problem promptly provide appropriate instructions and training to plant operators on how to recognize the problem if it occurs and take appropriate mitigating actions, (3) to request that licensees notify the NRC concerning the existence of the problem at their facility, and (4) to request that licensees inform the NRC of measures taken to correct this and/or other significant problems that are identified as a result of this bulletin.

Description of Circumstances:

There have been four recent cases where a design deficiency has been found involving the minimum flow recirculation paths for ECCS pumps. Although these four cases all involve safety injection (SI) pumps in Westinghouse-designed  ;

reactors, similar problems also could exist in other systems and at other types of reactors. This design deficiency was first discovered at the Point Beach Nuclear Plant and was subseauently described in IE Information Notice 85-94 A similar problen involving residual heat removal (RHR) system loop selection logic was later found in several BWR plants. This problem was addressed in IE Compliance Bulletin 86-01. ,

On July 24, 1985 Wisconsin Electric Company submitted a report in accordance with 10 CFR Part 21 for the Point Beach Nuclear Plant describing a design i defic tency involving the minimum flow recirculation valves for the SI pumps.

At Point Beach the discharge lines for each of the SI pumps are connected to a cocinon recirculation header to provide a test flow path and a recirculation flow ,

path for minimum flow at times when the reactor coolant system (RCS) pressure exceeds the 51 pump shutoff head. The corrnon recirculation header is provided with two air-operated valves in series. These valves close to isolate the <

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1 October 8, 1986 l Page 2 of 3 refueling water storage tank (RWST from the containment sump during the recir-culation phase of emergency core co)oling following a postulated loss-of-coolant ,

accident (LOCA). Closure of-these' valves is' intended to prevent containment reactor coolant from being pumped outside containment to the RWST during the trecirculation phase. Both of the recirculation header isolation valves are 3 designed to fail closed when their control circuits lose electrical power or, control air pressure. The Part 21 report noted that a single failure (open) of _j the breaker associated with either of the two valves would isolate the minimum $

flow path for both z SI pumps, defeat the control room remote operation capability . i of the affected valve, and c'ause the loss of control room valve position j indication.

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On february 5,1986, Carolina Power and Light submitted LER 86-01 describing  ;

essentially the identical design deficiency involving the minimum flow recircu- I lation path for the SI pumps at H. B. Robinson. On June 20, 1986, Rochester Gas and Electric discovered a similar design deficiency at the Ginna Plant and on June 25, 1986, Florida Power and Light Company reported a similar design deficiency at the Turkey Point Plant. 3 The concern in all four cases above involves a postulated small break LOCA which initiates a safety injection signal that starts the SI pumps. During a small break LOCA, RCS pressure may not readily decrease below the SI pump shutoff head, A single failure resulting in the loss of the minimum flow path concurrent with 51 pump actuation would cause the pumps to operate deadheaded until RCS pressure j decayed below the SI pump shutoff head. The simultaneous loss of minimum flow valve position indication in the control room will exacerbate 'this' loss 6f'thinimum flow path. The availability of valve position indication is not expected to suf ficiently ameliorate this event. Operating the SI pumps deadheaded would result in pump damage and failure within a few minutes. The failure of multiple l

trains in an ECC5 due to a single failure violates the single failure criterion {

in General Design Criterion (GDC) 35 (10 CFR 50, Appendix A). In all the above l Cases, the short-term Corrective actions taken by the licensees were to mechani-  !

cally block open the SI pumo recirculation valves to ensure a minimum flow path  !

and to revise the applica % plant LOCA procedures to manually close these valves i prior to switching to the recirculation mode. This short term action should be carefully weighed against the requirements to minimize containment leak paths in the ECCS recirculation mode of operations and the reliability of operator actions in this regard.

Actions Required:'

1. Promptly determine whether or not your facility has a single-failure  !

vulnerability in the minimum flow recirculation line of any ECCS pumps  !

that could cause a failure of more than one ECCS train. .,

2. If the problem exists: (a) promptly instruct r11 operating shifts of the j problem and measures to recognize and mitigate the problem; (b) promptly '

develop and irnpleraent corrective actions which bring your facility into l compliance with GDC 35.

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• Actions required of the BWR plants in response to IE Compliance Bulletin 86-01 1 need not be repeated in responding to this Bulletin.

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Enclosu as 2 IEB 86-03 b October 8, 1986 Page 3 of 3

3. Within 30 days of receipt of this bulletin, (a) provide a written report to the NRC which identifies whether or not this problem exists at your f acility, (b) if the problem exists (or existed), include in the report the justification for continued operation and identify the short-term modifications to plant operating procedures or hardware that have been or -

are being implemented to ensure safe plant operations.

4. If the problem exists (or existed), prov.ide a written report within 90 days of receipt of this bulletin infoming the NRC of the schedule for long-tem resolution of this and/or any other significant problems that are identified as a result of this bulletin, j The written report shall be submitted to the appropriate Regional Administrator l

under oath or af firmation under provisions of Section 182a, Atomic Energy Act i l

of 1954, as amended. Also, the original copy of the cover letter and a copy of j the report shall be transmitted to the U.S. Nuclear Regulatory Commission, i Document Control Desk, Washington, D.C. 20555 for reproduction and distribution.  !

This request for information was approved by the Office of Hanagement and Budget under blanket clearance number 3150-0011. 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.

If you have questions regarding this matter, please contact the Regional ,

Administrator of the appropriate NRC regional of fice or one of the technical  ;

contacts. listed below. '

l

( p Jatnes Qw }d M.. Taylorgloirector s lf I

fifice of Inspection and Enforcement

• J Technical

Contact:

Henry Bailey, IE (301) 492-9006 Ron Young, IE (301) 492-8985 '

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Attachment:

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ENctosul2E 3 I  !

(J4ev1 /of3) 1 Wisconsin Elecinc ec..ia counur . I 231 W thCHlGAR P O COX 2046.t*LWAUKEE.WIL3201 (414)277 2345 VPNPD-86-465 NRC-86-110 t 1

1 i

November 12, 1986 1 Mr. J. G. Keppler, Regional Administrator Office of Inspection and Enforcement, Region III U. S. NUCLEAR REGULATORY COMMISSION ,

799 Roosevelt Road i Glen Ellyn, Illinois 60137 j 1

Dear Mr. Keppler:

DOCKETS 50-266 AND 50-301 l RESPONSE TO IE BULLETIN 86-0a POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2  ;

i This letter is in response to the request for information in IE i Compliance Bulletin 86-03.  ;

On July 24, 1986, Point Beach Nuclear plant notified the NRC, pursuant to the provisions of 10 CFR 21, of the potential defect in the design of the control circuit for the safety injection (SI) pump recirculation flow path isolation valves, i' SI-897Ar.B. It was discovered that a failure of the power ,

supply breaker from-the control-circuitrycofecither4S97. valve #

.would result-in closure of the valves as well asmloss.ofe annunciator and status light indication & intended:^toialert thee operator that an 897 valve was shuto Shutting of either 897 valve without operator knowledge or action could result in j failure of either or both SI pump (s) if pressure was above the l pump shutoff head.

]

At Point Beach Nuclear Plant, the SI pumps have a common return 1 header to the refueling water storage tank (RWST). This path provides a route for water flew during testing as well as a i recirculation path during pump operation while the reactor coolant system pressure is above the shutoff head of the pumps, a If the SI pumps are operated without any flow-through, overheating will damage the pump in a relatively short period of time. Immediate operator action is warranted in these l circumstances.

8611140171 G61112  :

PDR ADOCK 05000266 ,

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Mr. J. G. Keppler November 12, 1986 .

Page 2 l

)

Thec897 valves are interlocked with the containment. sump e l

, isolation valves, SI-851A&B.* This interlock prevents 4 contaminated containment sump water from being pumped into the RWST during the safety injection recirculation phase of a '

design basis accident.

As discussed in the 10 CFR 21 report of July 24, 1986, the short-term corrective actions were outlined. They included the  ;

following:

The manual handwheel operators on the 897 valves were .

positioned to override the automatic operators and .

maintain the valves in an open position. This  !

prevents closure of the 897 valves anytime either of j them would lose control power. It also prevents '

closure of the 897 valve whenever either of the 851 valves is open. The control room control switches and .

the local manual valve operators for the 851 valves, 1 which are normally closed, were locked shut. The '

Emergency Operating Procedures which control the '

switchover from the injection phase to the recirculation phase require the 897 valves to be closed prior to the opening of either 851 valve. An operations Special Order was also issued explaining '

the reason for the locks on the valves and the conditions under which they could be removed and the valves opened.

• The final corrective actions which have been installed are as follows:

i The manual handwheel operators on the 897_ valves wille  !

remain in the open position so that the automatic > i operators are overridden.s Locks on the local manual ,

operators of the 851 valves will remain installed.  !

' Electrical interlocks which1 replace the locks on the control room control switches ;have- been installed tog prevent-the opening of any 851 valve u'ntil at least, one 897 valve is fully closedi This assures that potentially highly contaminated sump water is not pumped-to the RWST. Additionally, the relay for the annunciator and status light has been changed from a normally deenergized relay with normally open contacts to a normally energized relay with normally closed contacts. The relay change ensures that, upon a loss of control power, proper status light indication and annunciator action occur.

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[,, . . . _ - _ _ _ _ . _ . .

, EAJCLOSUQG 3 (Shee13 &l3)

Mr. J. G. Keppler November 12, 1986  !

Page 3 A review of the low head safety injection systems at Point Beach Nuclear Plant has been done. We believe that no other potential for failure of multiple trains 1 of pumps due to single failure of valves in minimum .;

flow recirculation lines exists. The low head safety i injection system at Point Beach is the residual heat removal (RHR) system. Each train has its own recirculation line. If failure of a valve in one of the RHR train recirculation lines occurred, the other train would, therefore, not be affected. .

If you have any questions concerning the information provided, please contact my office.

Very truly yours, v-I.

Lld 6 C. W. Fay Vice Precident Nuclear Pvver C'eles to NRC Resident Inspector NRC Document Control Desk Washington, D. C. (with original)

Subscribed and sworn to before me this 12 122- day of /devemb r:v 1986. 3 a euo ih ,Uw vcs evde <

Notary Public, State'of Wisconsin .

My Commission expires 5-2'7- 9 0 .

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(s!wofIof5) d VdSCOMSin Elecinc rana coursur 231 W MICHIGAN.P O BOX 2046. MILWAUKEE.W43201 #414; ppt.p345 l l

VPNPD-88-606

^

10 CFR 50.73

NRC-88-126 5

i December 16, 1988 i

U. S. NUCLEAR REGULATORY COMMISSION l Document Control Desk

Mail Station Pl-137 Washington, D. C. 20555 l Gentlemen:

i DOCKET 50-266 AND 50-301 a ' LICENSEE EVENT REPORT 88-009-00 l

! 55 ADEQUACY OF EOP-1.3,

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l

" CONTAINMENT SUMP RECIRCULATION" l l POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 l i

i Enclosed is Licensee Event Report 88-009-00 for Point Beach Nuclear Plant, Units 1 and 2. This report is provided in ,

j i accordance with 10 CFR 50.73(a)(2)(v), "Any event or condition i that alone could have prevented the fulfillment of the safety  ;

function of structures or systems that are needed to...(B) Remove l l

residual heat..."

i

! This report details the finding of a procedural inadequacy such I

that insufficient time may be provided for manual actions prior to a

initiating operation of the residual heat removal system with pump

, suction from the containment sump during a large loss of coolant -

l accident. Corrective actions to resolve this potential problem l are also described in the report.

i

If any further information is needed, please do not hesitate to l contact us.

l l Very truly yours, l h <~

Am l i C. W. Fay i Vice resident l Nuclear Power Enclosure a

! Copies to URC Regional Administrator, Region III NRC Resident Inspector ygP I

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' L . t.# B Inadeauacy of EOP-1.3, " Transfer to Containment Sump Recirculation" l

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During an internal vertical slice audit of the residual heat ,

removal system initiated by Wisconsin Electric, an inadequacy was

, identified in an emergency operating procedure.

Emergency Operating Procedure (EOP) 1.3, " Transfer to Containment i Sump Recirculation," presently directs the operator to perform the  !

Iinal steps of establishing containment sump recirculation when  ! ,

the refuelit.g water storage tank (RWST) level decreases to 10 A percent. However, a caution in the procedure requires that all l pumps whien are taking suction from the RST be secured upon i reaching 6 percent level. Calculations made during the audit j which considered pump flow rates during a large break LOCA in flow i rates greater than those assumed during development of the EOp ,

concluded that sufficient time does not exist to complete the i ,

procedure steps prior to reaching 6 percent RWST level. This ,

could result in securing all safeguards pumps for a short period of time. Corrective actions have been recommended by the audit team, and a procedure revision is being pursued.

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0l0 0]2 0F 0l4 EVENT DESCRIPTION A vertical slide audit of the residual heat removal (RHR) system ,

was conducted in September and October 1988. The audit team i reviewed the operating procedural controls of the system during i its emergency functions, including the capability to provide l suction from the containment sump for core injection and containment spray flow. At 28 percent RWST level, EOp 1.3, ,

" Transfer to Containment Sump Recirculation," directs the operator to perform some preliminary steps in preparation for sump recirculation. At 10 percent RWST level, the procedure directs the operator to perform the final recirculation alignment. In a caution statement, the procedure also directs the operator to

, secure all pumps which are taking suction from the RWST when the RWST level drops to 6 percent. It was noted during the audit that the volume of water available from 10 percent to 6 percent RWST level would be depleted quickly during a large break LOCA. If no i operator action is taken, other than that specified in the i procedure, the following safeguards pumps would be taking suction from the RWST as level approaches 6 percent: two containment i spray pumps, one RHR pump, and one safety injection pump. If j normal pump flow rates for these conditions are used, 6 percent

! RWST level would be reached prior to completing the recirculation alignment for one safeguards train. Depending on how fast the manual actions specified by the procedure would be completed, calculations made during the audit concluded that safeguards flow i could be stopped for two to four minutes. ,

BACKGROUND This condition was identified during an internal vertical slice l audit conducted by Wisconsin Electric. The description of the finding, along with recommended corrective actions, was documented ,

in an audit finding report which was issued as part of the final ,

report. The audit finding was transmitted for response on  ;

i November 18, 1988.

GENERIC IMPLICATIONS No generic issues were identified as a result of the investigation  !

of this condition. -

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0l0 g;3 or g l4 tim,, . - - ~ ..c, mu.,nis REPORTABILITY This licensee event report is provided in accordance with 10 CFR 50.73(a)(2)(v), "Any event or condition that alone could have prevented the fulfillment of the safety of structure or systems that are needed to ...(B) remove residual heat." More specifically, paragraph 10 CFR 50.73(a)(2)(vi) states, " Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors..."

CAUSE A higher level of emphasis was placed on the more likely event of a small break LOCA when the EOP was written. This resulted in certain local actions being delayed until after 10 percent RWST level is reached. In conjunction with other specific actions in the procedure, this resulted in the EOP being deficient for a large b re a k LOCA .

SAFETY ASSESSMENT During the extremely unlikely event of a large break LOCA, this condition could have prevented fulfillment of a safety function if the operating crew was unaware of the potential difficulties with the procedure. In the worst case, this could have resulted in termination of safety injection flow to the core and termination of containment spray flow for a brief period of time (two to four minutes). ,

based on experiences in operator training, operating crews were aware of the problems that could be encountered with this procedure in the event of a large break LOCA. During simulator exercises, operators took compensatory actions during large break LOCA scenarios in anticipation of the timing problem with final recirculation alignment. Additionally, the probability of a double ended guillotine reactor coolant pipe break is extremely low. With these two facts considered, it is concluded that the actual safety impacts of this condition are minimal.

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0l0 0 l4 0F 0 l4 rwa- .- _ _~~c u m .,un CORRECTIVE ACTIONS Short-term An order will be issued to the operating personnel by December 31, 1988, which informs them of this potential problem during implementation of EOP 1.3. .

Long-te rm EOP 1.3 will be revised by June 1, 1989. The revision ,

will assure that sufficient time is available to perform the containment sump recirculation alignment under the worst cases for the loss of coolant accidents.

SIMILAR OCCURRENCES Y

l There have been no occurrences similar to this LER at Point Beach.

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,! 231 W MICHIGAN.P.O. BOX 2046. MILWAUKEE.W153201 (414)221 2345 l

! VPNPD-89-300 10 CFR 50.73 NRC-89-060 May 19, 1989 l

U. S. NUCLEAR REGULATORY COIC4ISSION Document Control Desk 1 Mail Station P1-137  :

Washington, D. C. 20555  ;

Gentlemen I

DOCKET 50-266 AND 50-301 l l LICENSE EVENT REPORT 89-004-00 j NONCONSERVATIVE ANALYSIS OF TRANSFER TO ,

CONTAINMENT SUMP RECIRCULATI7N j 4

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 1

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) Enclosed is Licensec Event Report 89-004-00 for Point Beach l Nuclear Plant, Units 1 and 2. This report is provided in j

! accordance with 10 CFR 50.73(a)(2)(v), "Any event or condition  ;

j that alone could have prevented the fulfillment of the safety ,

j function of structures or systems that are needed to...(B) Remove j residual heat..." . l 3 .

l i

This report details the finding of a previous analysis inadequacy l which impacts the time available for manual actions necessary to l l cffect the switchover of the residual heat removal and safety .

i injection systems from the injection phase to the containment sump i j recirculation phase while one system train is out of service and

during a large break LOCA. Corrective actions to resolve the ,

potential problem are also described in this report. -

j If any further information is needed, please do not hesitate to contact us. ,

l l .

j Very truly yours, l I d

• l

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C. W. Fay i Vice President i i Nuclear Power 8905300360 850519 '

F'DR ADOCK 05000266

Enclosure S FDC l

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On April 21, 1989 it was confirmed that the estimated time to core uncovery, assuming a single train of ECCS during the time of transfer from the refueling water storage tank to the containment sump, was ,

probably considerably shorter than previously assumed. The previous analyses had not considered steam entrainment and steam voiding in I the core. A similar issue had been discussed in LER 88-009-00; however, the cause discussed in that LER was due to an inadequate  ;

procedure. Calculations following LER 88-009-00 nad conservatively estimated that flow to the core would be stopped for approximately 7.5 minutes during the transfer using EOP 1.3, " Transfer to Con-tainment Sump Recirculation." The estimated time to core uncovery at that time was 10 to 13 minutes after ECCS flow was stopped.

Assuming steam entrainment, the estimated time to uncovery is 1.5 to 3 minutes. A night order was issued to provide guidance to the operating crews for an event of this type. A revision of EOP 1.3 is being evaluated.

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EVENT DESCRIPTION:

On April 19, 1989, the NRC notified Wisconsin Electric that Westing-house personnel had identified that calculations done for another facility had indicated that stopping flow to the core during a large  ;

break LOCA could lead to core dry out within 1.5 to 3 minutes. This i '

information was confirmed by the licensee on April 21, 1989 during a j teleconference with Westinghouse personnel. It should be noted that l a specific calculation by Westinghouse for Point Beach Nuclear Plant a >

has not been done. Previous calculations of ECCS flow interruption ,

during transfer to recirculation for the large break LOCA had not j >

taken into account steam entrainment and voiding in the core.

Without consideration of steam entrainment and voiding, our calcu-

lations had indicated that core uncovery could occur 10 to 13 minutes after the cessation of all ECCS flow. Also, calculations '

and estimates of valve manipulation times and procedure completion

times for the transfer from the refueling water storage tank (RWST) to the containment sump for the suction of the ECCS indicated a conservative transfer time of approximately 7.5 minutes. The transfer time observed in walk-throughs and simulator runs is in the 2 to 4 minute range. This switchover is controlled by Emergency '

Operating Procedure EOP 1.3, " Transfer to Containment Sump Recircu-lation." This procedure provides for the short duration shut down of all ECCS flow to the core, if the failure of one train of'ECCS  :

i occurs.

BACEGROUND:

i' The isolation of all ECCS flow guidance in EOP 1.3 was previously l identified as a concern during an internal vertical slice audit

, conducted by Wisconsin Electric. (See Point Beach Unit 1 LER 88-009-00) The discussion in this Licensee Event Report results from a calculation done by Westinghouse for another licensee. When the  !

] NRC became aware of the situation, Wisconsin Electric was notified.

3 Further discussions were held with the NSSS supplier that performed the calculation modeling steam entrainment and voiding to determine the applicability of the calculated time estimates to Point Beach Nuclear Plant. Specific calculations have not been completed for Point Beach.

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GENERIC APPLICATIONS:

Since at least one other licensee is affected by this analysis, it can be presumed that steam entrainment and steam voiding may not have been accounted for in the calculations for other licensees. The manual switchover from the RWST to the containment sump is an '

important consideration. At other plants, automatic switchover to RHR recirculation may result in little or no time during which all ECCS flow is stopped to the core.

REPORTABILITY:

This Licensee Event Report is provided in accordance with ,

10 CFR 50.73(a)(2)(v), "Any event or condition that alone could have prevented the fulfillment of the safety structures or systems that are needed to ... (B) remove residual heat." More e specifically, paragraph 10 CFR 50.73(a)(2)(Vi) states, " Events  ;

covered in paragraph (a)(2)(v) of this section may include one or '

more procedural errors,...or discovery of... analysis... inadequacies."  ;

. CAUSE:

I This EOP deficiency, dealing with the possible uncovery of the core, 1 appears to be due to the inadequate analysis of the interruption ,

of ECCS flow during a large break LOCA scenario. It was not recognized that the effects of phenomena such as steam entrainment '

and steam voiding in the core could be significant during the low RCS pressures resulting from a large break LOCA.

I' SAFETY ASSESSMENT:

3 During the extremely unlikely event of a large break LOCA with a i loss of one ECCS train, the procedural guidance could have stopped ECCS flow to the core if the operating crews were unaware of the e potential difficulties with the procedure. In the worst case, flow could have been stopped by way of safety injection for a brief time ,

(2 to 4 minutes).  ;

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Based on observed experiences in operator training, operating crews -

have demonstrated awareness of the concern related to the potential loss of flow to the core with the loss of one train of ECCS and the need to transfer from one water source to another during a large break LOCA. During the simulator exercises the operators have demonstrated anticipatory actions and compensatory actions to assure continued flow cf some kind to the core. Charging pumps can provide flow to the core during the transfer of ECCS suction from the RWST to the containment sump. The safety injection pump for the unaffected train could also be used to supply flow to the core during the transfer. Additionally, the probability of a double ended guillotine break in the reactor coolant pipe is extremely small. ,

Given the above, it is concluded that the safety impact of this condition is minimal.

CORRECTIVE ACTION:

• Short Term: On April 25, 1989, a night order was issued to the operating personnel discussing the scenario presented in this LER.  !

It was emphasized that core damage could occur within 90 seconds.

Two options were offered for the maintenance of ficv to the core assuming the loss of one train of ECCS. If instrument air is still  !

available, the use of charging pumps was suggested using both l normal and auxiliary charging lines. The continued operation of a  ;

safety injection pump with suction from the RWST was also listed as ,

an option while the same train residual heat removal pump is being i aligned to the containment sump.

Intern.ediate Corrective Action: The revision to EOP 1.3 will be pursued with the further consideration of the possibility of core dry out in a short time. Note that because of the additional considerations raised by this issue, our commitment in LER 88-009-00 i to provide a revision of EOP 1.3 by June 1, 1989 cannot be met.

The completion of this revision will occur by July 31, 1989.

Long Term: Further analysis and possible plant modifications must ,

be reviewed prior to the completion of any final EOP procedural or hardware modification corrective actions. Valves which are now manually operated may have to be modified for remote operation and as a consequence may have to be moved. The scope of these modifications is not known, but it is estimated that modifications, if necessary, would occur before or during the refueling outage for each unit in 1991.

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UCENSEE EVENT REPORT (LER) TEXT CONTINUATION amovse ows no siso-e* f sxm s. s<si,ss ,t i

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i SIMILAR OCCURRENCES: ,

There have been no occurrences similar to those described in this  !

LER at Point Beach. , l I

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