Information Notice 1998-22, Deficiencies Identified During NRC Design Inspections

Deficiencies Identified During NRC Design Inspections

ML031050142
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, Three Mile Island, 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, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	06/17/1998
From:	Roe J Office of Nuclear Reactor Regulation
To:
References
IN-98-022, NUDOCS 9806110395
Download: ML031050142 (9)

UNITES STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 June 17, 1998 NRC INFORMATION NOTICE 98-22: DEFICIENCIES IDENTIFIED DURING NRC DESIGN

INSPECTIONS

Addressees

All holders of operating licenses for nuclear power reactors, except those licensees who have

permanently ceased operations and have certified that fuel has been permanently removed

from the vessel.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this Information notice to alert

addressees to Issues identified during recent design team inspections regarding the capability

of selected systems to perform their design bases safety functions. It is expected that

recipients will review the information for applicability to their facilities and consider actions, as

appropriate, to avoid similar problems. However, suggestions contained in this Information

notice are not NRC requirements; therefore, no specific action or written response is required.

Background

In October 1996, as a result of concerns that some plant configurations and operations were

inconsistent with their design and licensing bases, the NRC formed three NRC-led teams of five

contracted engineers from architect-engineer firms to perform design-focused, inspections of

risk-significant safety systems. These Inspections were implemented to evaluate the capability

of the selected systems to perform their safety functions, the adherence of the selected

systems to their design and licensing bases, and the consistency of the as-built configuration

and system operation against the final safety analysis report (FSAR). The inspections were

conducted in accordance with NRC Inspection Procedure 93801, eSafety System Functional

Inspection,* and focused on the engineering design and configuration control sections of the

procedure. As of May 1, 1998, 16 Inspections have been completed.

Description of Circumstances

The following summarizes the most significant technical and programmatic issues that have

been identified by the 16 Inspections completed to date. In some Instances, there erg ongoing

NRC staff reviews concerning specific regulatory and technical aspects of the issues.

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IN 98-22 June 17,1998 Modifications or Evaluations That Resulted In Operation of the Plant Outside the

Licensing Bases

At some plants, there were Issues concerning operation outside the licensing bases

stated In the FSAR. At some plants, licensee 10 CFR 50.59 evaluations failed to Identify

unreviewed safety questions, and licensees made significant changes to plant

operations or equipment without NRC's approval. In some Instances, this was due to

revisions to calculations that were not subject to 10 CFR 50.59 screening programs. At

Farley and Perry, Inspectors Identified Issues involving the lack of protection from

tornado missiles. At Perry, the suppression pool cleanup (SPCU) system was

frequently operated at power, which was Inconsistent with the updated safety analysis

report (USAR) description. Due to the system Interface between the SPCU and the high

pressure core spray (HPCS) system, whenever the SPCU system Is In operation, the

HPCS system must be aligned to the suppression pool as opposed to its preferred

source, the condensate storage tank. At Diablo Canyon, the team identified a single- failure vulnerability with the component cooling water (CCW), auxiliary saltwater, and

residual heat removal (RHR) systems and an Inability to use the containment spray

system during the containment recirculation mode of the RHR system.

At Cooper, the licensee made a modification to the reactor equipment cooling system

which resulted in a leakage of 200 gallons per day from sampling valves that were

Inadvertently left In the open position. During the modification process, the need for

controlling the position of these valves was not recognized. Therefore, in accident

conditions, this leakage could have depleted the available water to the system, resulting

in an Inability of the system to support Its long-term post-accident cooling functions.

Errors In Analyses for Emergency Core Cooling System (ECCS) Pump Suction Swap- over from Refueling Water Storage Tank (RWST)/Borated Water Storage Tank (BWST)

to the Reactor Sump During a Loss of Coolant Accident (LOCA)

At five plants, licensees made errors in the calculations that were performed to ensure

adequate coolant would be available to support operation of low-pressure pumps during

and after swap-over from the RWST/BWST to the reactor building sump during a

postulated LOCA. These errors resulted from the use of non-conservative reactor

building pressures, valve stroke times, and operator response times; failing to account

for limiting system configurations; and Instrument uncertainties. The errors affected the

calculations for setting the RWST/BWST level Instrument alarms, the emergency

operating procedures (EOPs), and technical specifications. At Three Mile Island (TMII),

the licensee declared both the decay heat removal system and the reactor building

spray pumps Inoperable as a result of these concerns.

At D.C. Cook, the licensee did not sufficiently evaluate the Instrument uncertainties and

flow biases that would cause the RWST level Instrumentation to Indicate lower than

IN 98-22 June 17, 1998 actual level. In addition, the licensee was unable to demonstrate the adequacy of

drainage paths from the Inactive to the active containment recirculation sumps to

support operation of the ECCS pumps with suction from the ECCS sump. On the basis

of these concerns, the licensee declared both trains of the ECCS and the containment

spray system inoperable and Initiated a dual-unit shutdown.

At H. B. Robinson, a design modification by the licensee allowing for single safety

Injection (SI) pump operation resulted In insufficient net positive suction head (NPSH) for

two of three SI pumps. At Ginna, the licensee implemented procedure changes after a

slight negative NPSH was calculated for the RHR pump A.

At Wolf Creek, the licensee's RWST Instrument loop uncertainty calculations did not

consider density variations resulting from temperature changes and boron

concentration, which affected the RWST alarm and swap-over setpoints and the

accuracy of the RWST level Indication.

Inadequate Testing of Safety-Related Components

Inspection teams Identified numerous examples of Inadequate testing of safety related

components, including the lack of testing for certain molded-case circuit breakers at

St. Lucie, Arkansas Nuclear One (ANO), and TMI; leak and functional testing of valves

(including check valves) at TMI, St. Lucie, Ginna, Farley, D.C. Cook, Palisades, and

Davis-Besse; post-modification testing of safety injection pumps at H.B Robinson;

testing of the safety Injection lock-out relay at Indian Point 2; and testing of sections of

the auxiliary service water supply path and pumps at Diablo Canyon. In some

instances, the licensees did not perform periodic tests, while In other cases, the testing

was inadequate to demonstrate the operability of all safety functions. For example, the

licensee tested certain check valves In the forward flow but not the reverse flow

direction. At Vermont Yankee, the inspectors determined that testing of the RHR heat

exchangers was Inadequate because of invalid test instrument uncertainty assumptions.

Issues Concerning Implementation of Computer Evaluation Models Used for Analyzing

ECCS Response to Design Basis Accidents

In three inspections, the team Identified Issues concerning the computer evaluation

models used for analyzing the ECCS response to postulated design-bases accidents.

At H.B. Robinson, the computer model Indicated the existence of a second peak in fuel

clad temperature that was significantly higher than the peak temperature reported by the

licensee. At Ginna, the team Identified errors In the analysis report that called Into

question the level of review and the validity of some Inputs. At Indian Point 2, the

licensee had not established procedural controls to ensure that Input data assumptions

used In the model would not be Invalidated by plant modifications. The team also

Identified the lack of formal design control procedures between the licensee and the

vendor for supplying and verifying the validity of Input data and assumptions.

IN 98-22 June 17, 1998 System Operation at a Temperature In Excess of the Design Basis

At three plants, the inspection team found that the plant had been operated while

system ambient temperatures were inexcess of the design or licensing bases. At

Vermont Yankee, the licensee aflowed the suppression pool temperature to exceed the

design-basis temperature used inthe analyses for the standby cooling system pump

NPSH, containment pressure, piping stress, and equipment qualification. At D.C. Cook, the licensee operated the plant with the essential service water temperature (ultimate

heat sink) Inexcess of the design-basis temperature, which could affect the qualified life

of equipment Inthe control room and reduce the rate of heat removal from the spent fuel

pool.

At Palisades, LOCA analyses concluded that the post-accident CCW temperatures

could exceed the design-basis temperature; however, the licensee had not evaluated

system performance at the higher temperatures.

Errors Made in Evaluating Post-accident Temperatures for Safety Related Pump Rooms

At four plants, licensees made errors in calculating the maximum pump room

temperatures that would be expected during post-accident conditions. At Palisades, Ginna, and Indian Point 2,the calculations for the auxiliary feedwater pump rooms

incorrectly used nominal rather than the maximum expected ambient temperature

conditions before the accident. At Cooper, the licensee calculated the heat load for the

RHR pump room using only one of the two pumps. Also at Cooper, operating

procedures were Inadequate to ensure that the maximum calculated RHR service water

booster pump room temperature would not be exceeded.

Lack of Controls or Specified Outage Times for Limiting System Line-ups That Couild

Challenge Design-Basis Considerations

Inspectors identified issues pertaining to the lack of controls or specified outage times

(either within technical specifications or administrative procedures) for ensuring systems

are maintained Ina configuration that would support all design basis considerations. At

Cooper, the licensee had no controls in place to limit the time the RHR system is

operated Inthe suppression pool cooling mode. Inthis mode of operation, the system

would not be capable of automatically realigning Into the Injection mode given certain

single-failure assumptions.

At Indian Point 2, Inspectors found that there were no controls for taking Instrument

busses off their invertors and supplying them from alternate power sources. Under

accident conditions, certain Instrumentation would be lost because of the shedding of

loads from the alternate sources. Also at Indian Point 2, Inspectors found that there

were no controls for limiting the time the Si system could be used to fill accumulator

tanks. During this process, Ifa safety systems actuation occurred, a portion of the Si

flow credited Inthe accident analysis would be diverted.

IN 98-22 June 17, 1998 Other Significant Issues Identified During the Course of the Inspections:

At Vermont Yankee, the licensee operated the RHR pumps at minimum flow values that

were significantly less than those recommended by the pump vendor.

At Robinson, control cables for all three Si pumps were routed in the same conduit In

violation of single failure and separation criteria.

At ANO, vendor-specified flow limits for the steam generators were not incorporated Into

plant procedures. As a result, operators were unaware that flow limits were exceeded

during a plant transient.

At D.C. Cook and Cooper, failures of instrument air regulators could result in the

Inoperability of redundant safety trains as a result of the over pressurization of air

operated valves.

Discussion

The majority of the issues identified have resulted from errors In the original design or design

modifications, calculational errors, inadequate corrective action, inadequate testing, and

documentation discrepancies. Many of the original design, design modifications, and

calculational errors can be attributed to the Inadequate specification and control of system and

discipline Interfaces, inadequately verified calculational assumptions, or the use of superseded

calculations. Licensees failed to evaluate the impact of calculational revisions on other

calculations, operating and test procedures. Changes to operating and test procedures were

not always reviewed against the existing calculations to ensure calculational assumptions were

still bounding. Also, the lack of a controlled, easily retrievable design basis has, in some

instances, hindered the ability of licensee engineers to Identify all design basis safety functions

of a system or component

Inadequate corrective actions have often resulted from weaknesses In root cause analyses or

from failing to assign ownership to engineering Issues. In addition, the depth of Internal self

assessments has not always been sufficient to Identify configuration management

weaknesses.

Additional details regarding the specific issues Identified during the NRC design Inspections can

be found In the following NRC Inspection Reports:

K> IN 98-22 June 17, 1998 PLANT INSPECTION REPORT# ACCESSION #

Ginna 50-244/97-201 9710060295 H.B. Robinson 2 50-261/97-201 9708280104 Vermont Yankee 50-271/97-201 9709020247 Diablo Canyon 50-275197-202 9712030153 Three Mile Island 50-289196-201 9704210100

Arkansas Nuclear 1 50-313197-201 9803120197 Donald C. Cook I & 2 50-315/97-201 9712030232 St. Lucle 1 & 2 50-335196-201 9703280271

9703280234 Davis-Besse 50-346197-201 9709180174 Joseph M. Farley I & 2 50-348/97-201 9705230286 Washington Nuclear 2 50-397/96-201 9704250204 Indian Point 2 50-247/98-201 9804020083 Palisades 50-255/97-201 9801130395 Wolf Creek 50-482/97-201 9803030253 Cooper 50-298/97-201 9802260235 Perry 1 50-440/97-201 9706130253 This Information notice requires no specific action or written response. If you have any

questions about the Information In this notice, please contact one of the technical contacts lIsted

below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Jeffrey Jacobson, NRR

301-415-2977 E-mall: jbj@nrc.gov

Attachment: Ust of Recently Issued NRC Information Notices 74iP6-e  % fiAY

DOCUMENT NAME: S:MDRPM SEC\98-22.lN Tech Ed. concurred on 5/15/98 To receive a copy of this document, Indicate In the box C-Copy wbo ettachment/enclosure ECopy wfth attachmentlenciosure N v No copy

lOFFICE PECB* l PECB* l PECB* l C:PECBl (A)D:DRPM

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DATE 06102/98 06/08/98 06/08198 06/09/98 06/ 0198

• See previous concurrenceOFFICIALRECORDCOPY

Attachment

IN 98-22 June 17, 1998 Page 1 of I

LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

98-21 Potential Deficiency of 6/4198 All holders of operating licenses

Electrical Cable/Connection for nuclear power reactors, except

Systems those licensees who have

permanently ceased opertions

and have certified that fuel has

been permanently removed from

the reactor vessel.

98-20 Problems with Emergency 613/98 All holders of operating licenses

preparedness Respiratory for nuclear power reactor;

Protection Programs non-power reactors; all

fuel cycle and material licensees

require to have an NRC approval

emergency plan

98-19 Shaft Binding In General Electric 6/3/98 All holders of operating licenses

Type SBM Control Switches for nuclear power reactors

98-18 Recent Contamination Incidences 5/13/98 Part 35 Medical Ucensees

Resulting from Failure to Perform

Adequate Surveys

98-17 Federal Bureau of Investigations 517/98 All U.S. Nuclear Regulatory

(FBI) Awareness of National Commission fuel cycle and power

Security Issues and Responses and non-power reactor licensees

(ANSIR) Program

98-16 Inadequate Operational Checks 4130/98 All Industrial Radiography

of Alarm Ratemeters Licensees

98-15 Integrity of Operator Licensing 4120/98 All holder of operating licenses

Examinations for nuclear power reactors except

those that have permanently

ceased operations and have

certified that fuel has been

permanently removed from the

reactor vessel

OL = Operating License

CP = Construction Permit

IN98-22 June 17, 1998 PLANT INSPECTION REPORT # ACCESSIONA

Ginna 50-244/97-201 9710060295 H.B. Robinson 2 50-261/97-201 9708280104 Vermont Yankee 50-271/97-201 9709020247 Diablo Canyon 50-275/97-202 9712030153 Three Mile Island 50-289/96-201 9704210100

Arkansas Nuclear 1 50-313/97-201 9803120197 Donald C. Cook 1 & 2 50-315/97-201 9712030232 St. Lucie 1 & 2 50-335/96-201 9703280271

9703280234F

Davis-Besse 50-346/97-201 9709180174 Joseph M. Farley 1 & 2 50-348197-201 9705230286 Washington Nuclear 2 50-397/96-201 970425020n4 Indian Point 2 50-247/98-201 9804020083 Palisades 50-255/97-201 9801130395 Wolf Creek 50-482/97-201 9803030253 Cooper 50-298/97-201 9802260235 Perry 1 50-440/97-201 9706130253 This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact one of the technical contacts listed

below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

JT . IXoeoyActing Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Jeffrey Jacobson, NRR

301-415-2977 E-mail: jbj~nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: S:%DRPMSEC\98-22.IN Tech Ed. concurredwith on 5/15198 flnrfirnfn

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OFFICE PECB* I PECB* I PECB* I C:PEC (A)D:DRP_

NAME JJacobson TKoshy RDennig JStolz JRoe

DATE 06/02/98 06108/98 j 06/08/98 J_06/09/98 . 06/10/98 OFFICIAL RECORD COPY

IN 98-XX

/ June xx, 1998 PLANT INSPECTION REPORT# ACCESSION##

.Ginna 50-244/97-201 9710060295 H.B. Robinson 2 50-261/97-201 9708280104 Vermont Yankee 50-271/97-201 9709020247 Diablo Canyon 50-275/97-202 9712030153 Three Mile Island .50-289/96-201 9704210100

Arkansas Nuclear 1 50-313/97-201 9803120197 Donald C. Cook 1 & 2 50-315/97-201 9712030232 St. Lucie 1 & 2 50-335/96-201 9703280271

9703280234 Davis-Besse 50-346/97-201 9709180174 Joseph M. Farley 1 & 2 50-348/97-201 9705230286 Washington Nuclear 2 50-397/96-201 9704250204 Indian Point 2 50-247/98-201 9804020083 Palisades 50-255/97-201 9801130395 Wolf Creek 50-482/97-201 9803030253 Cooper 50-298/97-201 9802260235 Perry 1 50-440/97-201 9706130253 This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact one of the technical contacts listed

below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

Jack W. Roe, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Jeffrey Jacobson, NRR

301-415-2977 E-mail: jbj~nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\TXK\AEINSPI.

To receive a co of this document, indicate in the box C=Copy wo attachmentlenclosure E=Copy with attachmentlenclosure N = No copy

OFFICE PECB l PECB l PEC4% I l) C B l (A lP l

NAME JJacobson TKoshy J JRX

DATE 06hC198 06h5N98 06/'t,98 066/ l0I98 OFFICIAL RECORD COPY