United States Nuclear Regulatory Commission - Protecting People and the Environment


ACCESSION #:  9806300529

                       LICENSEE EVENT REPORT (LER)



FACILITY NAME:  Nine Mile Point Unit 2                    PAGE: 1 OF 6



DOCKET NUMBER:  05000410



TITLE:  Systems Outside the Design Basis Due to Incorrect Valve

        Weights



EVENT DATE:  05/25/98   LER #:  98-014-00   REPORT DATE:  06/24/98



OTHER FACILITIES INVOLVED:                          DOCKET NO:  05000



OPERATING MODE:  5   POWER LEVEL:  000



THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR

SECTION:

50.73(a)(2)(i) & 50.73(a)(2)(ii)



LICENSEE CONTACT FOR THIS LER:

NAME:  Ray Dean, Engineering Manager -      TELEPHONE:  (315) 349-4240

       NMP2



COMPONENT FAILURE DESCRIPTION:

CAUSE:      SYSTEM:       COMPONENT:       MANUFACTURER:

REPORTABLE TO EPIX:



SUPPLEMENTAL REPORT EXPECTED:  NO



ABSTRACT:



On May 25, 1998, Nine Mile Point Unit 2 (NMP2) determined that

differences between actual valve weights and weights shown on engineering

drawings could have caused pipe stresses to exceed design allowables on

four piping systems.  These systems included High Pressure Core Spray

(CSH), Residual Heat Removal (RHS), Reactor Core Isolation Cooling

(RCIC), and Reactor Floor Drains (DFR).  NMP2 was shut down in Refueling

Outage 6 (RFO6) with the reactor cavity flooded and the core off loaded

at the time of discovery.  The systems were determined to be operable or

not required for the current plant shutdown conditions with the exception

of RHS Loop C.  RHS Loop C was already removed from service for

outage-related activities.



The root cause of this event was failure of the vendor to provide

accurate valve weights during initial construction.  The actual valve

weights were not consistent with the vendor supplied drawings.



The valve drawings and the associated calculations were revised.

Engineering Supporting Analyses were performed to determine operability.

A review of other small bore valves was performed.  Piping configuration

changes were made such that design requirements were reestablished.  The

procurement process has been revised to require verification of small

bore valve weights during receipt inspection.



END OF ABSTRACT







TEXT                                                          PAGE 2 OF 6



I.  DESCRIPTION OF EVENT



On May 25, 1998, Nine Mile Point Unit 2 (NMP2) determined that

differences between actual valve weights and weights shown on engineering

drawings could have caused pipe stresses to exceed design allowables on

four piping systems.  These systems included High Pressure Core Spray

(CSH), Residual Heat Removal (RHS), Reactor Core Isolation Cooling

(RCIC), and Reactor Floor Drains (DFR).  NMP2 was shut down in Refueling

Outage 6 (RFO6) with the reactor cavity flooded and the core off loaded

at the time of discovery.  The systems were determined to be operable or

not required for the current plant shutdown conditions with the exception

of RHS Loop C.  RHS Loop C was already removed from service for

outage-related activities.



In 1997, NMP2 personnel identified a discrepancy with the valve weights

of small bore ASME Class 2 and 3 manual valves.  The identified valves

were all safety-related.  The actual weights of 522 valves were

determined to be higher than the weights shown on the vendor valve

drawings by as much as 50 percent.  The use of incorrect valve weight

impacts pipe stresses, pipe support/tie-back support loads and

qualification of valve accelerations.  Since the valves were all manual

valves, and the pipes and pipe supports/tie-back supports are passive

components, the safety functions for these components consisted of

maintaining structural integrity and thus the pressure boundary.



An Engineering Supporting Analysis (ESA) was performed using a sampling

of various calculations that may have been impacted.  Based on

calculations for 400 valves that were reviewed for a variety of

locations, loading conditions and configurations, and the conservatism

included in the calculations, the affected valves, piping and systems

were determined to meet design requirements.



The affected calculations were reviewed over a period of time to document

the qualifications of the impacted piping with the new valve weights.  On

May 25, 1998, after the documentation of the affected calculations was

completed, it was determined that of the 522 valves affected, a total of

eight valves on four different systems caused the piping on those systems

to not meet design requirements under normal operating and accident

conditions.  These affected systems included CSH, RHS, RCIC, and DFR.  An

additional ESA was performed which determined that the piping for six of

the eight valves associated with the CSH, RCIC and DFR Systems were

operable for the current plant shutdown conditions.  The two remaining

valves associated with RHS rendered RHS Loop C inoperable for the current

plant conditions.  However, RHS Loop C was already removed from service

for outage-related activities.



Conservatisms used in the calculations were re-evaluated, and of the

eight valves identified on May 25, 1998, three valves were determined to

meet design requirements under all conditions.  This included both valves

associated with the CSH System and therefore, the CSH System was

unaffected and capable of meeting its required functions at all times.

Thus, the piping associated with only five valves did not meet design

requirements for all plant conditions.  These five valves are described

further in the Analysis of Event section of this LER.



TEXT                                                          PAGE 3 OF 6



II.  CAUSE OF EVENT



The root cause of this event was failure of the vendor to provide

accurate valve weights during initial construction.  The actual valve

weights were not consistent with the vendor supplied drawings.



III.  ANALYSIS OF EVENT



This event is reportable in accordance with 10CFR50.73(a)(2)(i)(B), "Any

operation or condition prohibited by the plant's Technical

Specifications," and 10CFR50.73(a)(2)(ii), "Any event or condition that

resulted in the condition of the nuclear power plant, including its

principal safety barriers, being seriously degraded; or that resulted in

the nuclear power plant being: (B) In a condition that was outside the

design basis of the plant."



Technical Specification (TS) 3/4.4.8, "Structural Integrity," requires

that the reactor coolant system structural integrity of ASME Code Class

1, 2, and 3 components be maintained.  When the integrity of these

components fails to meet the applicable requirements, the affected

components must be returned to within limits or must be isolated.  The

piping associated with the two RHS valves (part of the reactor coolant

pressure boundary) did not meet the applicable requirements since

original installation due to the described deficiency, and since this

condition was not recognized, the applicable TS actions were not taken.

In addition, the applicable TS actions for system inoperability were not

taken and other systems which were required to be operable as a result

may not have been operable.



The impact of the affected valves and systems not meeting design

requirements is described below:



RHS Valves 2RHS*V220 and V221



The RHS System is designed to remove decay and sensible heat during and

after plant shutdown, inject water into the Reactor Pressure Vessel (RPV)

following a Loss of Coolant Accident (LOCA) to reflood the core

independently of other core cooling systems, and remove heat from the

primary containment following a LOCA, to limit the increase in primary

containment pressure and temperature.



Valves 2RHS*V220 and V221 are normally closed vent valves on a

three-quarter inch line on the RHS Loop C injection line.  These valves

are used as high point vents and also as a vent during Type C testing of

the Containment Isolation Valves (CIVs).  Assuming a three-quarter inch

hole on the injection line during a postulated LOCA, and assuming the

loss of Division I electrical power (single active failure), the RHS Loop

C injection capacity would have been slightly reduced.  However, such a

small reduction would not have significantly affected the heat removal

and core cooling function because injection flow rates used in the



TEXT                                                          PAGE 4 OF 6



III. ANALYSIS OF EVENT (Cont'd)



LOCA analysis are lower than current system performance test acceptance

criteria.  If a single active failure of the outboard CIV is assumed

(i.e., post-LOCA CIV will not close), the leakage through the hole would

have been confined within the RHS injection line boundary.  Any leakage

from the RHS line boundary (i.e., valve stem leakage) to the secondary

containment would have been treated by the Standby Gas Treatment System.

Therefore, changes to radiological consequences would likely have been

minimal.



RCIC Valve 2ICS*V225



The RCIC System provides adequate core cooling in the event the reactor

is isolated from its primary heat sink and feedwater flow is not

available.



Valve 2ICS*V225 is normally closed and isolates a one-half inch test

connection on the RCIC turbine exhaust line to the suppression pool,

which is used for Type C testing of the CIV.  For transients that would

initiate RCIC, steam would be released from the turbine exhaust line

which could lead to a RCIC isolation on area high temperature, assuming

that the one-half inch connection is broken.  This condition is alarmed

in the control room and thus the operators would have taken the

appropriate actions to place the plant in a safe condition.  CSH serves

as a backup to RCIC, can perform the same function as RCIC and would not

have been affected by the failure in RCIC, therefore assuring the ability

to place the plant in a safe condition.



The RCIC turbine exhaust line CIV is normally open.  For a postulated

LOCA, primary containment isolation would have been met even if the test

line was to fail.  The suppression pool water seal would have prevented

contaminated air leakage.  However, a small amount of water leakage would

be expected through the opening.  The consequences of such leakage are

small and would likely have been bounded by current radiological

analyses.  Local radiation alarm and/or flooding signals would have

alerted the operator to take corrective actions.



DFR Valves 2DFR*V112 and V113



The Reactor Floor Drains collect influent from radioactive or potentially

radioactive sources and high conductivity or potentially high

conductivity sources and discharge these fluids to the Radwaste System

for processing.



Valves 2DFRV*112 and V113 are normally closed and isolate a three-quarter

inch test connection on the floor drains leaving containment.  The valves

are used for Type C testing of the corresponding CIVs.  This pipe

connection is located in the air space above the suppression pool.  The

drain header is open to drywell atmosphere.  For the worst case scenario,

a suppression pool bypass path could have existed during a postulated

LOCA.  This bypass path would have resulted in an additional bypass area

of approximately six percent of design.  However, this additional bypass

area is within the Technical Specification (TS) Limiting Condition for

Operation (LCO) 3.6.2.1.b limit, which is 10 percent of the design.

Additionally, past



TEXT                                                          PAGE 5 OF 6



III.  ANALYSIS OF EVENT (Cont'd)



suppression pool bypass performance tests have shown that the actual

bypass area is approximately one percent of the design.  Therefore, the

event is within the existing limit and the containment barrier would have

been assured.



During a seismic event or LOCA, the pipe stress allowables of ASME

Section III Appendix F could have been exceeded and piping failures may

have occurred.  Although the above systems were determined to be outside

the design basis due to the incorrect valve weights and resultant pipe

stresses, the evaluations performed by NMP2 show that the systems would

have remained functional during normal plant operation.  As described

above, there was adequate protection for the reactor and containment

based on either redundant equipment or systems, or the minimal impact of

the failures on the associated systems.  Therefore, there were no adverse

consequences to the health and safety of the general public or plant

personnel.



IV.  CORRECTIVE ACTION



1.   The valve drawings were revised to correct the valve weights.



2.   The associated calculations were revised to reflect the actual valve

weights and ESAs were performed to determine operability.



3.   Piping configuration changes were made such that design requirements

were reestablished for the affected valves.  The changes included

reworking the weld contour, relocating or redesigning tie-back supports,

or removing valves and installing pipe caps.



4.   A review of other small bore valves supplied by this vendor was

performed.  Discrepancies were identified with two other valve sizes used

in the plant where indicated and actual valve weights were outside an

acceptable range.  In one case, the valves were lighter than shown on the

drawings and thus had no adverse impact.  In the other case, the valves

were heavier than shown on the drawings.  These configurations were

qualified analytically by reevaluating conservatisms used in the

calculations and thus were determined to meet design requirements.



5.   A requirement to verify small bore valve weights during receipt

inspection has been added to the procurement process.



V.   ADDITIONAL INFORMATION



A.   Failed components: none.



B.  Previous similar events: none.



TEXT                                                          PAGE 6 OF 6



V.   ADDITIONAL INFORMATION (Cont'd)



C.   Identification of components referred to in this LER:



     COMPONENT                IEEE 803 FUNCTION        IEEE 805 SYSTEM ID



Residual Heat Removal System            N/A                 BO



High Pressure Core Spray System         N/A                 BG



Reactor Core Isolation Cooling System   N/A                 BN



Reactor Floor Drain System              N/A                 WK



Valves                                  V              DO, BG, BN, WK



*** END OF DOCUMENT ***





Page Last Reviewed/Updated Thursday, March 29, 2012