IR 05000280/1986042
| ML18150A042 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/16/1987 |
| From: | Delmedico NRC |
| To: | Rehm T NRC |
| Shared Package | |
| ML18150A040 | List: |
| References | |
| NUDOCS 8704270037 | |
| Preceding documents: |
|
| Download: ML18150A042 (20) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555
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ONE HUNDREDTH, CONGRESS (
PHILIP R SHARP, INDIANA. CHAIRMAN e
ftOOM H2-331 HOUSE OFFICE BUILDING ANNEX 110 2 l'HONE 12021 221-noo 00\\JG WALGREN, PENNSYLVANIA CARLOS J. MOORHEAD, CALIFORNIA Al SWIFT, WASHINGTON WILLIAM E. DANNEMEYER. CALIFORNIA lt.6. J,oust of l\\tprtstntatibts MIKE SYNAR. OKLAHOMA JACK FIELDS. TE)(AS W.J. "BIU.Y TAUZIN, LOUISIANA MICHAEL G. OXLEY. OHIO IIILL IUCHARDSON. NEW MEXICO MICHAEL BILIRAKIS, FLORIDA Committrr on entrlll' anb Commmr JOHN BRYANT, TEXAS DAN SCHAEFER. COLORADO TERRY BRUCE. IWNOIS JOE BARTON, TEXAS EDWARD J. MARKEY, SONNY CALLAHAN, ALABAMA MASSACHUSETTS NORMAN F. LENT, NEW YORK SUBCOMMITTEE ON ENERGY AND POWER MICKEY UL.AND. TEXAS
!EX OFFICIOI IION WYDEN, OREGON IIAl.l'H M. HAU.. TEXAS WAYNE DOWDY, MISSISSIPPI Rl"1Jfngton, 13C 20515 JOHN D. DINGEU. MICHIGAN
!EX OFFICIO)
Mr. Walter Mikesell Robert L. Cloud and Associates 125 University Avenue Berkeley, CA 94710
Dear Mr. Mikesell:
March 16, 1987 The Subcommittee on Energy and Power is investigating the implications for the safety of nuclear power plants of the recent Surry accident.
In particu-lar, we are concerned that (1) despite the designation of the failed feedwater line as "a nonsafety related system," a similar failure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) ~tandards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
We are requesting your response to the following questions:
1. The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration of the feedwater suction line between the condenser and the feedwater pump.
(a) What codes, standards, specifications and regulatory requirements are applied to the failed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine, pipelines and components)? Are these systems classified as nuclear or non-nuclear? Are they classified as safety or nonsafety related systems?
(b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel? If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant accident? What is the safety justification for the differences?
- J
--~- **
,,
.,
e Mr. Walter Mikesell-2-March 16, 1987 (c) If a failure in the feedwater piping occurred at a similar location, e.g., bet*ween the condenser and feedwater piping in a Boiling Water Reactor nuclear power plant, could radioactive material be released outside the containment?
(i) If so, how much could be released and what would be the consequences to the surrounding area?
(ii) How are these areas of the feedwater and steam lines classified in Boiling Water Reactors?
(iii) In view of the Surry accident, do you think that the classifica-tions of these areas of the power plant (including the steam turbine, condenser and feedwater pumps) are appropriate?
(d) What additional requirements could be applied to the feedwater lines, steam lines, steam turbine, feedwater pumps, condenser and related equip-ment to improve the safety of nuclear plant operation?
(e) Do you think the NRC should make any changes in its regulatory require-ments for Surry or other nuclear power plants in order to implement lessons learned from the Surry accident?
2.
The NRC team reports cited erosion/corrosion induc~d thinning of pipe metal as the cause of the failure at the Surry Station.
Do the design, construction, maintenance or integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately prov1de for finding or make allowances for deterioration of plant compon~nts and piping in service? If not, what regulatory changes should the NRC make to incorporate these factors in plant design, inspection and maintenance requirements?
3. The two Surry Station nuclear units are very similar in design, nuclear reactor system and age.
The units also "share" some support and auxiliary functions.
(a) In view of this dependency, does it seem appropriate that Unit 1 was not shut down immediately when the failure occurred in Unit 2?
(b) Should the NRC issue any new regulatory guidance fo_r such situations?
4.
Changes in the control room ventilation system were being implemented while the plant was running and at the time of the accident.
The NRC inspection team reports conclude that the modification work resulted in the control room being flooded with potentially lethal carbon dioxide gas.
- I
".
Mr. Walter Mikesell-3-March 16, 1987 (a) Are NRC regulations adequate for modifications being performed while plants are operating?
Were these regulations being observed at the time of the accident?
(b) Do you feel that different procedures should have been used?
Should the NRC make any regulatory changes to prevent ongoing modification work from compromising operational safety?
5.
The NRC inspection team reports indicate the accident was initiated by an improperly maintained valve.
(a) Does it seem appropriate that the plant was allowed to operate with this valve not functioning properly? Are there adequate requirements for inspections of such valves?
(b) Should the NRC make any regulatory changes as a result of the mainte-nance deficiencies discovered during the investigation of this accident?
6.
What actions independent of NRC regulatory requirements should and has the industry taken to implement lessons learned from the Surry accident?
Thank you for your assistance with this investigation.
In reviewing your responses we will assume that you are drawing on your industry experience and associations.
However, we will consider your answers as your own personal opinions.
We would appreciate having your response no later than April 10.
PRS:bh I
ONE ~UNDREDTH CONGRESS *
PHILIP R. SHARP. INDIANA. CHAIRMAN I
e ROOM H2-331 HOUSE OFFICE BUILDING ANk[l IIO 2 l'tlONE (202) 226-2500
- . DOUG W"1.GREN. PENNSYLVo\\NIA AL SWIFT, WASHINGTON CARLOS J. MOORHEAD. CALIFORNIA WILLIAM E. DANNEMEYER. CALIFORNIA JACK FIELDS, TEXAS It.&. J,oust of l\\tprtstntatibts MIKE SYNAR. OKi>.HOMA W.J. -111LLY-- TAUZIN. LOUISIANA BILL RICHARDSON. NEW MEXICO JOHN BRYANT, TEXAS MICHAEL G. OXLEY. OHIO MICHAEL BILIRAKIS, FLORIDA DAN SCHAEFER, COLORADO Committtt on entrll!' anb Commmt TERRY BRUCE. IWNOIS JOE BARTON, TEXAS EDWARD J. MARKEY, SONNY CALLAHo\\N, ALABAMA NORMAN F. LENT, NEW YORK MASSACHUsms SUBCOMMITTEE ON ENERGY AND POWER MICKEY LELAND, TEXAS (IX OFFICIO)
IION WYDEN, OREGON RALPH M. HALL. TEXAS WAYNE DOWDY. MISSISSIPPI JOHN D. DINGELL. MICHIGAN RIHbington, 13C 20515 (EX OFFICIO)
Mr. Robert D. Pollard Union of Concerned Scientists 1616 P Street, NW Suite 310 Washington, DC 20036
Dear Mr. Pollard:
March 16, 1987 The Subcommittee on Energy and Power is investigating the implications for the safety of nuclear power plants of the recent Surry accident.
In particu-lar, we are concerned that (1) despite the designation of the failed feedwater line as "a nonsafety related system, 11 a similar failure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) standards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
We are requesting your response to the following questions:
1. The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration of the feedwater suction line between the condenser and the feedwater pump.
(a) What codes, standards, specifications and regulatory requirements are applied to the failed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine, pipelines and components)1 Are these systems classified as nuclear or non-nuclear1 Are they classified as safety or nonsafety related systems1 (b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel1 If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant accident1 What is the safety justification for the differences1
,,
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..
Mr. Robert D. Pollard-2-March. 16, 1987 (c) If a failure in the feedwater piping occurred at a similar location, e.g., between the condenser and feedwater piping in a Boiling Water Reactor nuclear power plant, could radioactive material be released outside the containment?
(i) If so, how much could be released and what would be the consequences to the surrounding area?
( ii) How are these areas of the f eedwater and steam lines classified in Boiling Water Reactors?
(iii) In view of the Surry accident, do you think that the classifica-tions of these areas of the power plant (including the steam turbine, condenser and feedwater pumps) are appropriate?
(d) What additional requirements could be applied to the feedwater lines, steam lines, st~am turbine, feedwater pumps, condenser and related equip-ment to improve the safety of nuclear plant operation?
(e) Do you think the NRC should make any changes in its regulatory require-ments for Surry or other nuclear power plants in order to implement lessons learned from the Surry accident?
2.
The NRC team reports cited erosion/corrosion induced thinning of pipe metal as the cause of the failure at the Surry Station.
Do the design, construction, maintenance or integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately provide for finding or make allowances for deterioration of plant components and piping in service? If not, what regulatory changes should the NRC make to incorporate these factors in plant design, inspection and maintenance requirements?
3.
The two Surry Station nuclear units are very similar in design, nuclear reactor system and age.
The units also 11share 11 some support and auxiliary functions.
(a) In view of this dependency, does it seem appropriate that Unit 1 was not shut down imme4iately when the failure occurred in Unit 2?
(b) Should the NRC issue any new regulatory guidance for such situations?
4.
Changes in the control room ventilation system were being implemented while the plant was running and at the time of the accident.
The NRC inspection team reports conclude that the modification work resulted in the control room being flooded with potentially lethal carbon dioxide gas.
vr. Robert D. Pollard-3-March 16, 1987 (a) Are NRC regulations adequate for modifications being performed while plants are operating? Were these regulations being observed at the time of the accident?
(b) Do you feel that different procedures should have been used?
Should the NRC make any regulatory changes to prevent ongoing modification work from compromising operational safety?
5.
The NRC inspection team reports indicate the accident was initiated by an improperly maintained valve.
(a) Does it seem appropriate that the plant was allowed to operate with this valve not functioning properly? Are there adequate requirements for inspections of such valves?
(b) Should the NRC make any regulatory changes as a result of the maintenance d~ficiencies discovered during the investigation of this accident?
6.
What actions independent of NRC regulatory requirements should the industry take to implement lessons learned from the Surry accident?
Thank you for your assistance with this investigation.
We would appreciate having your response no later than April 10.
PRS :bh
- J ONE HUNDREDTH CONGRESS
... *..
PHILIP R. SHARP, INDIANA. CHAIRMAN I
IIOOM H2-ll I HOUSE OFFICE BUILDING Ali"U JOO 2 l'HONE 12021 22&-2500 DOUG WALGREN. PENNSYLVANIA Al SWIFT, WASHINGTON CARLOS J. MOORHEAD. CAUFOIINIA WILLIAM E. DANNEMEYER. CALIFORNIA JACK FIELDS, TEXAS 11.&. J,oust of l\\tprtsentatibes MIKE SYNAR. OKLAHOMA W.J. "'BILL r TAUZIN, LOUISIANA IIILL IIICHARDSON. NEW MEXICO JOHN IIRY ANT, TEXAS MICHAEL G. OXLEY, OHIO MICHAEL BILIRAKIS, FLORIDA DAN SCHAEFER. COLORADO
<ommitttt on fntrllP anb <ommmt ltRIIY BRUCE. IWNOIS EDWARD J. MARKEY, JOE BARTON, TEXAS SONNY CALLAHAN, ALABAMA NORMAN F. LENT, NEW YORK MASSACHUSITTS SUBCOMMITTEE ON ENERGY AND POWER Rla~bington, llC 20515 MICltEY LELAND, TEXAS (EX OFFICIOI IION WYDEN. OREGON IIAll'H M. tw.l. TEXAS WAYNE DOWDY, MISSISSIPl'I JOHN D. DiNGEU. MICHIGAN (Et OfflCIDI Mr. Zach Pale President INFO 1100 Circle 75 Parkway Suite 1500.
Atlanta, GA 30339
Dear Mr. Pale:
March 16, 1987 The Subcommittee on Energy and Power is investigating the implications for the safety of nuclear power plants of the recent Surry accident.
In particu-lar, we are concerned that (1) despite the designation of the failed feedwater line as "a nonsafety related system," a similar failure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) standards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
We are requesting your response to the following questions:
1. The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration of the feedwater suction line between the condenser and the feedwater pump.
(a) What codes, standards, specifications and regulatory requirements are applied to the failed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine, pipelines and components)? Are these systems classified as nuclear or non-nuclear? Are they classified as safety or nonsafety related systems?
(b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel? If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant accident?
What is the safety justification for the differences?
'1
"4*
Mr. Zach Pale-2-March 16, 1987 (c) If a failure in the feedwater piping occurred at a similar location, e.g., between the condenser and feedwater piping in a Boiling Water Reactor
. nuclear power plant, could radioactive material be released outside the containment?
(i) If so, how much could be released and what would be the consequences to the surrounding area?
(ii) How are these areas of the feedwater and steam lines classified in Boiling Water Reactors?
(iii) In view of the Surry accident, do you think that the classifica-tions of these areas of the power plant (including the steam turbine, condenser and feedwater pumps) are appropriate?
(d) What additional requirements could be applied to the feedwater lines, steam -lines, steam turbine, feedwater pumps, condenser and related equip-ment to improve the safety of nuclear plant operation?
(e) Do you think the. NRC should make any changes in its regulatory require-ments for Surry or other nuclear power plants in *order to implement lessons learned from the Surry accident?
2.
The NRC team reports cited erosion/corrosion induced thinning of pipe metal as the cause of the failure at the Surry Station.
Do the design, construction, maintenance or integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately provide for finding or make allowances for deterioration of plant components and piping in service? If not, what regulatory changes should the NRC make to incorporate these factors in plant design, inspection and maintenance requirements?
3.
The two Surry Station nuclear units are very similar in design, nuclear reactor system and age.
The units also "share" some support and auxiliary functions.
(a) In view of this dependency, does it seem appropriate that Unit 1 was not shut down immediately when the failure occurred in Unit 2?
(b) Should the NRC issue any new regulatory guidance for such situations?
4.
Changes in the control room ventilation system were being implemented while the plant was running and at the* time of the accident. The NRC inspection team reports conclude that the modification work resulted in the control room being flooded with potentially lethal carbon dioxide gas.
I.
p
...
Mr. Zach Pale-3-March 16, 1987 (a) Are NRC regulations adequate for modifications being performed while plants are operating? Were these regulations being observed at the time of the accident?
(b) Do you feel that different procedures should have been used?
Should the NRC make any regulatory changes to prevent ongoing modification work from compromising operational safety?
S.
The NRC inspection team reports indicate the accident was initiated by an improperly maintained valve.
(a) Does it seem appropriate that the plant was allowed to operate with this valve not functioning properly? Are there adequate requirements for inspections of such valves?
(b) Should the NRC make any regulatory changes as a result of the maintenance deficiencies discovered during the investigation of this accident?
6.
What actions independent of NRC regulatory requirements should the industry take to implement lessons learned from the Surry accident?
Thank you for your assistance with this investigation.
We would appreciate having your response no later than April 10.
PRS :bh
ONE HUNDREDTH CONGRESS
..
'i
- "'Pl<'1u, R. SHARP, INDIANA. CHAIRMAN
-~'*
ftOOM H2-331 HOUSE OFFICE BUILDING ANNEX NO 2 P'l<ONE 12021 22&-2~00 DOUG WALGREN, 'fNNSYLVANIA CARLOS J. MOORHEAD. CALIFORNIA WILLIAM E. DANNEMEYER. CALIFORNIA JACK FIELDS. TEXAS AL SWIFT, WASHINGTON It&. J,ousc of l\\cprcscntatibcs MIKE SYNIAR. OKLAHOMA W.J. "81LLY-TAUZIN, LOUISIANA alLL IIICHARDSON, NEW MEXICO JOHN IIRYANT, TEXAS MICHAEL G. OXLEY, OHIO MICHAEL BILIRAKIS, FLORIDA DAN SCHAEFER. COLORADO (ommitttt on enrrll!' anb (ommmt TERRY llftUCE. IWNOIS JOE BARTON. TEXAS EOWARO J. MARKEY, SONNY CALLAHAN, ALABAMA NORMAN F. LENT, NEW YORK MASSACHUSETTS MICKEY LELAND, TEXAS IION WYOEN, OREGON IIAll'H M. HALL. TEXAS (EX OFFICIO)
SUBCOMMITTEE ON ENERGY AND POWER WAYNE DOWDY, MISSISSll'l'I Rlasbfngton, me 20515 JOHN O. DINGELL MICHIGAN (EX OfFICIO)
Mr. Donald F. Landers Teledyne Engineering Services 130 Second Avenue Waltham, MA 02254
Dear Mr. Landers:
March 16, 1987 The Subconunittee on Energy and Power is investigating the implications for the safety of nuclear power plants of the recent Surry accident.
In particu-lar, we are concerned that (1) despite the designation of the failed feedwater line as "a nonsafety related system," a similar failure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) standards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
We are requesting your response to the following questions:
1.
The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry
- Station was caused by service induced deterioration of the feedwater suction line between the condenser and the feedwater pump.
Ca) What codes, standards, specifications and regulatory requirements are applied to the failed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine., pipelines and components)? Are these systems classified as nuclear or non-nuclear? Are they classified as safety or nonsafety related systems?
(b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel? If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant
~ccident? What is the safety justification for the differences?
j f'
Mr. Donald F. Landers-3-March 16, 1987 (a) Are NRC regulations adequate for modifications being performed while plants are operating? Were these regulations being observed at the time of the accident?
(b) Do you feel that different procedures should have been used?
Should the NRC make any regulatory changes to prevent ongoing modification work from compromising operational safety?
5.
The NRC inspection team reports indicate the accident was initiated by an improperly maintained valve.
(a) Does it seem appropriate that the plant was allowed to operate with this valve not functioning properly? Are there adequate requirements for inspections of such valves?
(b) Should the NRC make any regulatory changes as a result of the mainte-nance deficiencies discovered during the investigation of this accident?
6. What actions independent of NRC regulatory requirements should and has the industry taken to implement lessons learned from the Surry accident?
Thank you for your assistance with this investigation.
In reviewing your responses we will assume that you are drawing on your industry experience and associations. However, we will consider your answers as your own personal opinions.
We would appreciate having your response no later than April 10.
y, -R,d~
~~~
Chairman PRS:bh
ONE HU,.OREOTI-I CONGRESS f
. --
'
PHILIP R. SHARP, INDIANA. CHAIRMAN
~DOM H2-J31 HOUSE OFFICE BUILDING ANNEX NO 2 l'HONE 12021 228-HOO i)
DOUG WAI.GREN. ~ENNSYlVANIA Al SWIFT. WASHINGTON CARLOS J. MOORHEAD. CALIFORNIA WlLllAM E. DANNEMEYER. CALIFORNIA JACK FIELDS, TEXAS
11.&,. J,ousc of ~cprcscntatibcs MIKE SYNAR. OKLAHOMA W.J. -lllllY"" TAUZIN. LOUISIANA lllll RICHARDSON. NEW MEXICO JOHN BRYANT. TEXAS MICHAEL G. OXLEY. OHIO MICHAEL BlllRAKIS, FLORIDA DAN SCHAEFER. COLORADO CommiUtt on entrin, anb Commtrce TERRY IIRUCE. llUNOIS JOE BARTON. TEXAS SONNYCALLAHAN.ALAIIAMA NORMAN F. LENT, NEW YORK EDWARD J. MARKEY, UASSACHUSms MICKEY LELAND. TEXAS ltX OFFICIO)
SUBCOMMITTEE ON ENERGY AND POWER 110N WYDEN. OREGON IIAU'H M. HALL.. TEXAS W.\\YNE DOWDY, MISSISSll'l'I JOHN D. DINGELL. MICHIGAN (VC OFFICIO)
Mr. Don~ld McDonald Executive Director National Board of Boiler and Pressure Vessel Inspectors 1055 Crupper Avenue Columbus, OH 43229
Dear Mr. McDonald:
March 16, 1987 The Subcommittee on Energy and Power is investigating the implications for the safety of nuclear power plants of the recent Surry accident. In particu-lar, we are concerned that (1) despite the designation of the failed feedwater line as "a nonsafety related system, 11 a similar failure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) standards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
We are requesting your response to the following questions:
1.
The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration of* the feedwater suction line between the condenser and the feedwater pump.
(a) What codes, standards, specifications and regulatory requirements are applied to the failed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine, pipelines and components)? Are these systems classified as nuclear or non-nuclear? Are they classified as safety or nonsafety related systems?
(b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel? If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant accident?
What is the safety justification for the differences?
~~-
.
. ',.
.,.r.
- Mr. Donald McDonald-2-March 16, 1987 (c) If a failure in the feedwater piping occurred at a similar location, e.g., between the condenser and feedwater piping in a Boiling Water Reactor nuclear power plant, could radioactive material be released outside the containment?
(i) If so, how much could be released and what would be the consequences to the surrounding area?
(ii) How are these areas of the feedwater and steam lines classified in Boiling Water Reactors?
(iii) In view of the Surry accident, do you think that the classifica-tions of these areas of the power plant (including the steam turbine, condenser and feedwater pumps) are appropriate?
(d) What additional requirements could be applied to the feedwater lines, steam lines, steam turbine, feedwater pumps, condenser and related equip-ment to improve the safety of nuclear plant operation?
(e) Do you think the NRC should make any changes in its regulatory require-ments for Surry or other nuclear power plants in order to implement lessons learned from the Surry accident?
2.
The NRC team reports cited erosion/corrosion induced thinning of pipe metal as the cause of the failure at the Surry Station.
Do the design, construction, maintenance or integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately provide for finding or make allowances for deterioration of plant components and piping in service? If not, what regulatory changes should the NRC make to incorporate these factors in plant design, inspection and maintenance requirements?
3.
The two Surry Station nuclear units are very similar in design, nuclear reactor system and age.
The units also "share" some support and auxiliary functions.
(a) In view of this dependency, does it seem appropriate that Unit 1 was not shut down immediately when the failure occurred in Unit 2?
(b) Should the NRC issue any new regulatory guidance for such situations?
4.
Changes in the control room ventilation system were being implemented while the plant was running and at the time of the accident.
The NRC inspection team reports conclude that the modification work resulted in the control room being flooded with potentially lethal carbon dioxide gas.
,..
Mr. Donald McDonald-3-March 16, 1987 (a) Are NRC regulations adequate for modifications being performed while plants are operating? Were these regulations being observed at the time of the accident?
(b) Do you feel that different procedures should have been used?
Should the NRC make any regulatory changes to prevent ongoing modification work from, compromising operational safety?
S.
The NRC inspection team reports indicate the accident was initiated by an improperly maintained valve.
(a) Does it seem appropriate that the this valve not functioning properly?
inspections of such valves?
plant was allowed to operate with Are there adequate requirements for (b) Should the NRC make any regulatory changes as a result of the mainte-nance deficiencies discovered during the investigation of this accident?
6.
What actions independent of NRC regulatory requirements should and has the industry taken to implement lessons learned from the Surry accident?
Thank you for your assistance with this investigation.
In reviewing your responses we will assume that you are drawing on your industry experience and associations.
However, we will consider your answers as your own personal opinions.
We would appreciate having your response no later than April 10.
ely,
,.._,~** ~&
R. Sharp Chairman
.
PRS :bh
ONE, HUNOREDTh'CONGRESS ROOM H2-lll
'.I,(
PHILIP R. SHARP. INDIANA. CHAIRMAN
HOUSE OFFICE BUILDING ANNEX ltO 2 PHONE (2021 22&-noo DOUG *:\\1t.1.G.RrN. PENNSYLVANIA CARLOS J. MOORf<EAD. CALIFORNIA AI...SWIFT. WASHINGTON WIUIAM E. DANNEMEYER. CALIFORNIA 11.6. Jl,oust of l\\tprtstntatibts MIKE SYNAR. OKLAHOMA JACK FIELDS. TEXAS W.J. -111u Y" TAUZIN, LOUISIANA MICf<AEL G. OXLEY, OHIO IIILL RICHARDSON. NEW MEXICO MICf<AEL BILIRAKIS, FLORIDA JOHN IIRYANT, TEXAS DAN SCHAEFER, COLORADO (ommittrt on ntrlU' anb (ommmt TERRY IIRUCE. IWNOIS JOE BARTON, TEXAS EDWARD J. MARKEY, SONNY CALLAHAN. ALABAMA MASSACHUSETTS NORMAN F. LENT, NEW YORK SUBCOMMITIEE ON ENERGY AND POWER MICKEY LELAND, TEXAS If)( OFFICIOI RON WYOEN, OREGON RALPH M. HAU. TEXAS WAYNE DOWDY, MISSISSIPPI JOHN 0. DINGELL MICHIGAN (EX OFFICIOI Dr. Spencer H. Bush Batelle Pacific Northwest Laboratories PO Box 999 Richland, WA 99352
Dear Dr. Bush:
March 16, 1987 The Subcommittee on Energy and Power is investigating the implications for the safety of nuclear power plants of the recent Surry accident.
In particu-lar, we are concerned that (1) despite the designation of the failed feedwater line as "a nonsafety related system, 11 a similar failure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) standards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
We are requesting your response to the following questions:
1. The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration of the feedwater suction line between the condenser and the feedwater pump.
(a) What codes, standards, specifications and regulatory requirements are applied to the failed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine, pipelines and components)? Are these systems classified as nuclear or non-nuclear? Are they classified as safety or nonsafety related systems?
(b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel? If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant accident?
What is the safety justification for the differences?
~--
Dr. Spencer H. Bush-2-March 16, 1987 (c) If a failure in the feedwater piping occurred at a similar location, e.g., between the condenser and feedwater piping in a Boiling Water Reactor nuclear power plant, could radioactive material be released outside the containment?
(i) If so, how much could be released and what would be the consequences to the surrounding area?
(ii) How are these areas of the f eedwater and steam lines classified in Boiling Water Reactors?
(iii) In view of the Surry accident, do you think that the classifica-tions of these areas of the power plant (including the steam turbine, condenser and feedwater pumps) are appropriate?
(d) What additional requirements could be applied to the feedwater lines, steam lines, steam turbine, feedwater pumps, condenser and related equip-ment to improve the safety of nuclear plant operation?
(e) Do you think the NRC should make any changes in its regulatory require-ments for Surry or other nuclear power plants in order to implement lessons learned from the Surry accident?
2.
The NRC team reports cited erosion/corrosion induced thinning of pipe metal as the cause of the failure at the Surry Station.
Do the design, construction, maintenance o.r integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately provide for finding or make allowances for deterioration of plant components and piping in service? If not, what regulatory changes should the NRC make to incorporate these factors in plant design, inspection and maintenance requirements?
3. The two Surry Station nuclear units are very similar in design, nuclear reactor system and age.
The units also "share" some support and auxiliary functions.
(a) In view of this dependency, does it seem appropriate that Unit 1 was not shut down immediately when the failure occurred in Unit 2?
(b) Should the NRC issue any new regulatory guidance for such situations?
4.
Changes in the control room ventilation system were being implemented while the plant was running and at the time of the accident.
The NRC inspection team reports conclude that the modification work resulted in the control room being flooded with potentially lethal carbon dioxide gas.
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Dr. Spencer H. Bush-3-March 16, 1987 (a) Are NRC regulations adequate for modifications being performed while plants are operating? Were these regulations being observed at.the time.of the accident?
(b) Do you feel that different procedures should have been used?
Should the NRC make any regulatory changes to prevent ongoing modification work from compromising operational safety?
S.
The NRC inspection team reports indicate the accident was initiated by an improperly maintained valve.
(a) Does it seem appropriate that the plant was allowed to operate with this valve not functioning properly? Are there adequate requirements for inspections of such valves?
(b) Should the NRC make any regulatory changes as a result of the mainte-nance deficiencies discovered during the investigation of this accident?
. 6.
What actions independent of NRC regulatory requirements should and has the industry taken to implement lessons learned from the Surry accident?
Thank you for your assistance with this investigation.
In reviewing your responses we will assume that you are drawing on your industry experience and associations.
However, we will consider your answers as your own personal opinions.
We would appreciate having your response no later than April 10.
PRS :bh
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ONE 0HUNDREDTH -CONGRESS PHILIP R. SHARP, INDIANA. CHAIRMAN *
ROOM H2-331 HOUSE OFFICE BUILDING ANNE.J< 110 2
!'HONE 12021 22&-2&00 DOUG WAI.GREN. PENNSYLVANIA Al SWIFT, WASHINGTON CARLOS J. MOORHEAD. CALIFORNIA WILLIAM E. DANNEMEYER. CALIFORNIA JACK FIELDS. TEXAS 11.&,. J,oust of l\\tprtstntatibts MIKE SYNAR. OKLAHOMA W.J. -BILLY-TAUZIN. LOUISIANA BILL RICHAROSON. NEW MEXICO JOHN BRYANT, TtXAS MICHAEL G. OXLEY, OHIO MICHAEL 91LIRAKIS, FLORIDA DAN SCHAEFER. COLORADO Committee on f nergp anb (ommme TERRY BRUCE. IWNOIS JOE BARTON. TEXAS EDWARD J. MARKEY, SONNY CAll..AHAN, ALABAMA NORMAN F. LENT. NEW YORK MASSACHUS£TTS SUBCOMMITTEE ON ENERGY AND POWER MICKEY LELAND, TEXAS l£X OFFICIO)
RON WYDEN, OREGON RALPH II. HAU. TEXAS WAYNE DOWDY. MISSISSIPPI JOHN D. DINGELL MICHIGAH
!EX OFflCIO)
Mr. Everett C. Rodabaugh E. C. Rodabaugh & Associates, Inc.
4625 Cemetery Road Hilliard, OR 46026
Dear Mr. Rodabaugh:
March 16, 1987 The Subcommittee on Energy and Power is investigating the implications for the safety of nuclear power plants of the recent Surry accident.
In particu-lar, we are concerned that (1) despite the designation of the failed feedwater line as "a nonsafety related system," a similar failure in a Boiling Water Reactor could result in the release of radioactive steam outside the contain-ment structure; and (2) standards established for new nuclear power plants and inspection procedures for operational plants may not adequately take into account the possibility of deterioration of materials.
_We are requesting your response to the following questions:
1. The NRC Augmented Inspection Team Reports Nos. 50-280/86-42 and 50-281/86-42 (NRC team reports) indicate that the failure at the Surry Station was caused by service induced deterioration of the feedwater suction line between the condenser and the feedwater pump.
(a) What codes, standards, specifications and regulatory requirements are applied to the failed feedwater line and associated equipment (condenser, feedwater pumps, steam turbine, pipelines and components)? Are these systems classified as nuclear or non-nuclear? Are they classified as safety or nonsafety related systems?
(b) Are these requirements different than those applicable to other por-tions of the feedwater and steam lines that are closer to the steam gen-erators and reactor vessel? If so, why are they, and do you think this distinction is appropriate in view of what occurred in the Surry Plant accident? What is the safety justification for the differences?
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Mr. Everett C. Rodabaugh-2-March 16, 1987 (c) If a failure in the feedwater piping occurred at a similar location, e.g., between the condenser and feedwater piping in a Boiling Water Reactor nuclear power plant, could radioactive material be released outside the containment?
(i) If so, how much could be released and what would be the consequences to the surrounding area?
(ii) How are these areas of the feedwater and steam lines classified in Boiling Water Reactors?
(iii) In view of the Surry accident, do you think that the classifica-tions of these areas of the power plant (including the steam turbine, condenser and feedwater pumps) are appropriate?
(d) What additional requirements could be applied to the feedwater lines, steam lines, steam turbine, feedwater pumps, condenser and related equip-ment to improve the safety of nuclear plant operation?
Ce) Do you think the NRC should make any changes in its regulatory require-ments for Surry or other nuclear power plants in order to implement lessons learned from the Surry accident?
2.
The NRC team reports cited erosion/corrosion induced thinning of pipe metal as the cause of the failure at the Surry Station.
Do the design, construction, maintenance or integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately provide for finding or make allowances for deterioration of plant components and piping in service? If not, what regulatory changes should the NRC make to incorporate these factors in plant design, inspection and maintenance requirements?
3. The two Surry Station nuclear units are very similar in design, nuclear reactor system and age. - The units also "share" some support and auxiliary functions.
(a) In view of this dependency, does it seem appropriate that Unit 1 was not shut down immediately when the failure occurred in Unit 2?
(b) Should the NRC issue any new regulatory guidance for such situations?
4.
Changes in the control room ventilation system were being implemented while the plant was running and at the time of the accident.
The NRC inspection team reports conclude that the modification work resulted in the control room being flooded with potentially lethal carbon _dioxide gas.