Information Notice 1997-09, Inadequate Main Steam Safety Valve (MSSV) Setpoints and Performance Issues Associated with Long MSSV Inlet Piping

Inadequate Main Steam Safety Valve (MSSV) Setpoints and Performance Issues Associated with Long MSSV Inlet Piping

ML031050370
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:	03/12/1997
From:	Martin T Office of Nuclear Reactor Regulation
To:
References
IN-97-009, NUDOCS 9703140143
Download: ML031050370 (9)

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UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 March 12, 1997 NRC INFORMATION NOTICE 97-09: INADEQUATE MAIN STEAM SAFETY VALVE

(MSSV) SETPOINTS AND PERFORMANCE ISSUES

ASSOCIATED WITH LONG MSSV INLET PIPING

Addressees

All holders of operating licenses or construction permits for nuclear power reactors.

Purpose

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

licensees to the recent staff findings related to improper main steam safeLy valve (MSSV)

setpoints and MSSV performance issues associated with long inlet piping. 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.

Description of Circumstances

The licensee for Millstone Unit 2, a Combustion Engineering pressurized-water reactor

(PWR), reviewed the calculations used to determine MSSV setpoints to assure that peak

main steamline pressure did not exceed the allowable pressure. The licensee determined

that the calculations may be inadequate because the dynamic pressure loss between the

main steamline and the MSSVs was not modeled. This stretch of piping is not modeled in

the licensee's design-basis event transient analysis. The consideration of the dynamic

pressure drop in the piping would reduce the relieving capacity of the MSSVs. Therefore, the

omission resulted in an underprediction of the peak main steamline pressure by a relatively

significant amount and also resulted in the potential for the calculated peak pressure to

exceed 110 percent of the design pressure for the main steam system. In addition, significant dynamic pressure drops resulting from long inlet piping could cause unstable

MSSV performance.

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IN 97-09 March 12, 1997 Discussion

The peak main steam pressure for PWRs generally occurs when a main steam isolation

valve closure or a turbine trip is postulated. Typically, as much as 100 percent of the full

main steam flow can be relieved through the MSSVs following one of these events. These

high flowrates can create significant dynamic (i.e., frictional and acoustic) pressure drops.

At Millstone Unit 2, 8 individual code safety relief valves are attached to each of the 2 main

steamlines with approximately 6.1 m [20 feet] of 15.24-cm [6-inch] piping. The full-flow

pressure drop through this 6.1 m [20 ft] of piping could be as high as 689 kPa [100 psi]. As

a result of not modeling the stretch of piping for each MSSV, the actual discharge capacity of

the MSSVs was overpredicted and the peak main steamline pressure was underpredicted.

The relatively long stretch (approximately 6.1 m [20 feet]) of relatively small (15.24-cm

(6-inch]) piping between the main steamlines and the MSSVs contributed to the magnitude of

the underprediction of the peak pressure; however, the dynamic pressure loss from all

stretches of piping should be accounted for in the analysis. For plants that do not have long

stretches of narrow piping or have a large manifold, the MSSVs will not have as large a

pressure drop, however, the calculations may still be affected significantly depending on the

actual piping configuration in the plant.

In addition, long MSSV inlet piping may also affect the stability of the MSSVs. If dynamic

pressure drops are great enough, the valve disks may chatter because (after the valves

open) the valve inlet pressures will immediately drop below the valve reseating pressures.

Because significant excess system pressure has not been relieved,.the valves reopen and

the chattering cycle would continue.

Depending on the specific plant, this analysis can be performed by the architect engineer, the

nuclear steam supply system vendor, the fuel vendor, or the licensee. For example, the fuel

vendor may supply the transient analysis; however, the system pressure losses that are

inputs to the transient analysis are frequently provided by the architect engineer. As a result, it is important that controls are in place such that data communication across organizational

interfaces include all pressure losses in the transient analysis.

IN 97-09 March 12,1997 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.

Thomas T. Martin, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Christopher P. Jackson, NRR Charles G. Hammer, NRR

(301) 415-2947 (301) 415-2791 E-mail: cpjenrc.gov E-mail: cghenrc.gov

Eric J. Benner, NRR

(301) 415-1171 E-mail: ejblnrc.gov

Attachment: List of Recently Issued NRC Information Notices

K>

Attachment

IN 97-09 March 12, 1997 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

97-08 Potential Failures 03/12/97 All holders of OLs

for General Electric or CPs for nuclear

Magne-Blast Circuit power reactors

Breaker Subcomponents

97-07 Problems Identified 03/06/97 All holders of OLs

During Generic Letter or CPs for nuclear

89-10 Closeout power reactors

inspections

97-06 Weaknesses in Plant- 03/04/97 All holders of OLs

Specific Emergency or CPs for nuclear

Operating Procedures power reactors with

for Refilling the with once-through

Secondary Side of Dry steam generators

Once-Through Steam

Generators

91-85, Potential Failures of 02/27/97 All holders of OLs

Rev. 1 Thermostatic Control or CPs for nuclear

Valves or Diesel power reactors

Generator Jacket

Cooling Water

97-05 Offsite Notification 02/27/97 All holders of OLs

Capabilities or CPs for nuclear

power reactors and

test and research

reactors

97-04 Implementation of a New 02/24/97 All materials, fuel

Constraint on Radioactive cycle, and non-power

Air Effluents reactor licensees

OL = Operating License

CP = Construction Permit

IN 97-09 March 12, 1997 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.

original signed by M.M. Slosson

Thomas T. Martin, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Christopher P. Jackson, NRR Charles G. Hammer, NRR

(301) 415-2947 (301) 415-2791 E-mail: cpj@nrc.gov E-mail: cghenrc.gov

Eric J. Benner, NRR

(301) 415-1171 E-mail: ejbl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:IEJB1\SRV.IN *SEE PREVIOUS CONCURRENCE

Tech Editor has concurred on 12/3/96 To receive a copy of this document, Indicate In the box: C"' u Copy without enclosures "E"- Copy with enclosures "N" u No copy

OFFICE

CONTACT

S I AC:C/SRXB C:PECB I D:DRPM I

NAME . CJackson* TCollins* AChaffee* TMartin

GHammer*~fC I

EBenner* _

DATE 12/9/10/11/96 12/12/96 01/16/97 103/t7/97 OFFICIA7L RECORD COPY

IN 97-XX

January , 1997 At Millstone Unit 2, 8 individual code safety relief valves are attached to each of the 2 main

steamlines with approximately 6.1 m [20 feet] of 15.24-cm [6-inch] piping. The full-flow

pressure drop through this 6.1 m [20 ft] of piping could be as high as 689 kPa [100 psi]. As

a result of not modeling the stretch of piping for each MSSV, the actual discharge capacity of

the MSSVs was overpredicted and the peak main steamline pressure was underpredicted.

The relatively long stretch (approximately 6.1 m [20 feet]) of relatively small (15.24-cm

[6-inch]) piping between the main steamlines and the MSSVs contributed to the magnitude of

the underprediction of the peak pressure; however, the dynamic pressure loss from all

stretches of piping should be accounted for in the analysis. For plants that do not have long

stretches of narrow piping or have a large manifold, the MSSVs will not have as large a

pressure drop, however, the calculations may still be affected significantly depending on the

actual piping configuration in the plant.

In addition, long MSSV inlet piping may also affect the stability of the MSSVs. If dynamic

pressure drops are great enough, the valve disks may chatter because (after the valves

open) the valve inlet pressures will immediately drop below the valve reseating pressures.

Because significant excess system pressure has not been relieved, the valves reopen and

the chattering cycle would continue.

Depending on the specific plant, this analysis can be performed by the architect engineer, the

nuclear steam supply system vendor, the fuel vendor, or the licensee. For example, the fuel

vendor may supply the transient analysis; however, the system pressure losses that are

inputs to the transient analysis are frequently provided by the architect engineer. As a result, it is important that controls are in place such that data communication across organizational

interfaces include all pressure losses in the transient analysis.

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.

Thomas T. Martin, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Christopher P. Jackson, NRR Charles G. Hammer, NRR

(301) 415-2947 (301) 415-2791 E-mail: cpjpnrc.gov E-mail: cgh@nrc.gov

Eric J. Benner, NRR

(301) 415-1171 E-mail: ejbl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\EJB1\SRV.IN *SEE PREVIOUS CONCURRENCE

To receive a copy of this document, Indicate In the box: "Vc- Copy without enclosures "E - Copy with enclosures "N" u No copy

OFFICE

CONTACT

S :CSRXB T:PECB D:DRPM AA

NAME CJackson* TCollins* AChaffee* TMart

Gl-ammer*\ I

EBenner* _ _97 _

DATE 12/9/10/11/96 12/12/96 01/16/97 XI/ /97 OFFICIAL RECORD COPY

M4WI :1/if

IN 97-XX

January , 1997 At Millstone Unit 2, 8 individual code safety relief valves are attached to each of the 2 main

steamlines with approximately 20 feet of 6-inch piping. The full-flow pressure drop

through this 20 feet of piping could be as high as 100 psi. As a result of not modeling the

stretch of piping for each MSSV, the actual discharge capacity of the MSSVs was

overpredicted and the peak main steamline pressure was underpredicted. The relatively long

stretch (approximately 20 feet) of relatively small (6-inch) piping between the main steamlines

and the MSSVs contributed to the magnitude of the underprediction of the peak pressure;

however, the dynamic pressure loss from all stretches of piping should be accounted for in

the analysis. For plants that do not have long stretches of narrow piping or have a large

manifold, the MSSVs will not have as large a pressure drop, however, the calculations rnay

still be affected significantly depending on the actual piping configuration in the plant.

In addition, long MSSV inlet piping may also affect the stability of the MSSVs. If dynamic

pressure drops are great enough, the valve disks may chatter because (after the valves

open) the valve inlet pressures will immediately drop below the valve reseating pressures.

Because significant excess system pressure has not been relieved, the valves reopen and

the chattering cycle would continue.

At Millstone Unit 2, the transient analysis is currently performed by Siemens. Depending on

the specific plant, this analysis can be performed by the architect engineer, the nuclear steam

supply system vendor, the fuel vendor, or the licensee. For example, the fuel vendor may

supply the transient analysis; however, the system pressure losses that are inputs to the

transient analysis are frequently provided by the architect engineer. As a result, i is

important that controls are in place such that data communication across organizational

interfaces include all pressure losses in the transient analysis.

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.

Thomas T. Martin, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Christopher P. Jackson, NRR Charles G. Hammer, NRR

(301) 415-2947 (301) 415-2791 E-mail: cpj@nrc.gov E-mail: cgh@nrc.gov

Eric J. Benner, NRR

(301) 415-1171 E-mail: ejbl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:XEJBISRV.IN *SEE PREVIOUS CONCURRENCE

To receive a copy of this document, Indicate In the box: *C u Copy without enclosures "E" a Copy with enclosures "N"

• No copy

OFFICE

CONTACT

S AC:C/SRXB I

C:PECBll Z D: DRPM

FI

NAME Cackson* TCollins* AChaff~e TMartin

GHammer*

EBenner* l _ _ _

DATE 12/9/10/11/96 12/12/96 /7 /96 12/ /96 OFFICIAL RECORD C / / 1

IN 96-XX

December XX, 1996 through this 20 feet of piping could be as high as 100 psi. As a result of not modeling the

stretch of piping for each MSSV, the actual discharge capacity of the MSSVs was

overpredicted and the peak main steamline pressure was underpredicted. The relatively long

stretch (approximately 20 feet) of relatively small (6-inch) piping between the main steamlines

and the MSSVs contributed to the magnitude of the underprediction of the peak pressure;

however, the dynamic pressure loss from all stretches of piping should be accounted for in

the analysis. For plants that do not have long stretches of narrow piping or have a large

manifold, the MSSVs will not have as large a pressure drop, however, the calculations may

still be affected significantly depending on the actual piping configuration in the plant.

In addition, long MSSV inlet piping may also affect the stability of the MSSVs. If dynamic

pressure drops are great enough, the valve disks may chatter because (after the valves

open) the valve inlet pressures will immediately drop below the valve reseating pressures.

Because significant excess system pressure has not been relieved, the valves reopen and

the chattering cycle would continue.

At Millstone Unit 2, the transient analysis is currently performed by Siemens. Depending on

the specific plant, this analysis can be performed by the architect engineer, the nuclear steam

supply system vendor, the fuel vendor, or the licensee. For example, the fuel vendor may

supply the transient analysis; however, the system pressure losses that are inputs to the

transient analysis are frequently provided by the architect engineer. As a result, it is

important that controls are in place such that data communication across organizational

interfaces include all pressure losses in the transient analysis.

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.

Thomas T. Martin, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Christopher P. Jackson, NRR Charles G. Hammer, NRR

(301) 415-2947 (301) 415-2791 E-mail: cpj@nrc.gov E-mail: cgh@nrc.gov

Eric J. Benner, NRR

(301) 415-1171 E-mail: ejbl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\EJB1\SRV.IN *SEE PREVIOUS CONCURRENCE

To receive a copy of this document, Indicate In the box: 'I' = Copy without enclosures "E" a Copy with enclosures "N" - No copy

OFFICE

CONTACT

S I AC:C/SRXB lC Z:PECB D:DRPM

NAME CJackson* TCollins* AChaffee TMartin

GHallmmer*

EBenner* _

DATE 12/9/10/11/96 12/12/96 112 /96 O /12/ /96 UVtMAL KtLUKU LUFT 'I\:

N

K..-' < IN 96-XX

December XX, 1996 each MSSV, the actual discharge capacity of the MSSVs was overpredicted and

the peak main steamline pressure was underpredicted. The relatively long

stretch (approximately 20 feet) of relatively small (less than 6-inch) piping

between the main steamlines and the MSSVs contributed to the magnitude of th

underprediction of the peak pressure, however, the dynamic pressure loss om

all stretches of piping should be accounted for in the analysis. For p nts

that do not have long stretches of narrow piping or have a large mani ld, the

MSSVs will not have as large a pressure drop, however, the calculatons may

still be affected significantly depending on the actual piping co iguration

in the plant.

In addition, long MSSV inlet piping may also affect the sta ity of the

MSSVs. If dynamic pressure drops are great enough, the v ye disk may chatter

because after the valves open, the valve pressure will i ediately drop below

the valve reseating pressures. Because significant ex ss system pressure has

not been relieved, the valves reopen and the chatteri g cycle would continue.

At Millstone Unit 2, the transient analysis is cur ntly performed by Siemens;

however, the responsibility for this type of ana sis is not always clearly

defined. Depending on the specific plant, this analysis can be performed by

the architect engineer, the nuclear steam sup y system vendor, the fuel

vendor, or the licensee. For example, the f el vendor may supply the

transient analysis; however, the system pr sure losses that are inputs to the

transient analysis are frequently provide by the architect engineer. As a

result, it is important to include all essure losses in the transient

analysis.

This information notice requires no pecific action or written response. If

you have any questions about the i formation in this notice, please contact

one of the technical contacts lised below or the appropriate Office of

Nuclear Reactor Regulation (NRR project manager.

Thomas T. Martin, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: C istopher P. Jackson. NRR Charles G. Hammer, NRR

01) 415-2947 (301) 415-2791 Internet: cpj@nrc.gov Internet: cgh@nrc.gov

Eric J. Benner, NRR

(301) 415-1171 Internet: ejbl@nrc.gov

Attachment: ist of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\ JB1\SRV.IN Q5 lul Ayz

To recelve a copy of th document, Indicate In the box: "C' - Copy without enclosures "E" - Copy with enclosures "N"u No copy

OFFICE CONT T AC:C/SRXB I C:PECB lI D:DRPM

NAME CJa son (e i TCollinn AChaffee TMartin

Rudessin GHminer t 4 ..kt/

-(-t )

lDATE r 12//961//

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