ML18150A039
| ML18150A039 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/17/1987 |
| From: | Zech L NRC COMMISSION (OCM) |
| To: | Sharp P HOUSE OF REP., ENERGY & COMMERCE |
| Shared Package | |
| ML18150A040 | List: |
| References | |
| NUDOCS 8704270015 | |
| Download: ML18150A039 (24) | |
Text
V Distribution:
UNITED STATES
.EDQ 2635
/
h,uCLEAR REGULATORY COMMISSION.. VStello
~0 ~ :2 yQ/ 2,. J".:f WASHINGTON, D. C. 20555 CHAIRMAN April 17, 1987 The Honorable Philip R. Sharp, Chairman Subcommittee on Energy and Power Committee on Energy and Commerce United States House of Representatives Washington, DC 20515
Dear Mr. Chairman:
JTaylor
~
1 TRehm EBeckjord JMurray JGrace EJordan DMi 11 er CPatel AJohnson PAD#2 Rdg.
CRossi TMurley/JSniezek Yo~r letter of March 16, 1987, requested information that would assist the Subcommittee on Energy and Power in their investigation of the implications of the safety of nuclear power plants in light of the recent ~urry accident.* Answers to the specific questions in your March 16, 1987, letter are enclosed.
The NRC continues to take an active interest in degradation of any nuclear power plant equipment which has relevance to the safety of the plant, and we will continue to monitor individual plant performance and overall industry experience in this area.
Where plant specific problems occur, the need for generic action will be assessed and appropriate corrective measures will be taken.
This is a normal part of the NRC'.s ongoing monitoring of industry performance.
In this regard the NRC staff will continue to assess the safety implications of the Surry feedwater pipe failure.
~I hope that the information provided will assist your review.
Commissioner Asselstine disagrees with this response and will provide his views in a separate letter.
Enclosure:
Answers to Specific Questions cc:
Rep. Carlos J. Moorhead Originated:
NRR:Miller Sincerely,
L
, Q~ESJION_l(a).
ANSWER.
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?
The requirements for the construction and inservice inspection of the safety-related systems differ from nonsafetv-related svstems because safety-related systems are relied upon to provide the capability to prevent or mitigate the consequences of accidents, remove heat from the reactor and maintain it in a safe shutdown condition.
The construction requirements of safety-related systems differ from nonsafety-related systems in the areas of materials inspection and non-destructive examination of piping system weldments, overpressure orotection, and quality assurance, including third party inspection.
For the main steam and feedwater system, the principal difference _between the design of the safety-related and nonsaf~ty-related components are that the safety-related systems are required to meet seismic criteria and requirements for design quality assurance which complies with 10 CFR 50, Appendix B.
Safety-related portions of these lines are also required to receive inservice inspection and testing under 10 CFR 50.55a(q), which invokesSection XI of the ASME Boiler and Pressure Vessel Code.
Nonsafety-related systems are not required by any NRC standard, or regulatory requirement to receive inservice inspection.
The term non-nuclear is not well defined, but as used by many and in the response below it describes piping not constructed to -
Section III of ASME Boiler and Pressure Vessel Code.
Power plants built prior to the adoption of Section III of the ASME Code were con~tructed* to other standards such as ANSI/ASME 831.1.
The condensate and feedwater systems of PWRs provide feedwater at the required temperature, pressure, and flow rate to the secondary side of the steam generators.
Condensate is pumped from the main condenser hotwell by the condensate pumps, passes through the low pressure feedwater heaters to the feedwater pumps, and then is oumped through the hiqh pressure feedwater heaters to the secondary side of the steam generators.
That portion of the condensate and feedwater system located within the turbine building and the portion of the feedwater lines between turbine building up to the containment isolation valves located outside the reactor containment building are not classified
/
rQUESION_l(a).
(Continued) safety-related.
The portion of the feedwater system from the containment isolation valves located outside the reactor containment buildin~ up to and including the secondary side of the steam generators are within the nuclear portion of the power plant and are classified safety-related.
An auxiliary feedwater system is connected to the main feedwater system and normally operates during startup, hot standby and shutdown to provide feedwater to the steam generators.
This system also functions as an emergency system for the removal of heat fro~ the primary system when the main feedwater system is not available and for emergency conditions including small loss-of-coolant accidents.
The entire auxiliary feedwater system is classified as a safety-related system.
Regulatory guidance with respect to the auxiliary feedwater system, the main feedwater system, main condensers and condensate system is provided in the following sections of Standard Review Plan, NUREG-0800, Revision 2, (July 1981) (Attached).
NUREG Section 10.4.1 10.4.7 1 0
- 4
- 9 Title Main Condensers Condensate and Feedwater System Auxiliary Fe~dwater System The following Regulatory Guides also provide guidance with respect to Quality Group Classification (applicable codes and standards), Seismic Design requirements, and Quality Assurance requirements for components of nuclear power plants (Attached).
Regulatory_Guide 1.26, Revision 3, (February 1976) 1.29, Revision 3, (Septemher 1978)
IJ..!l~
Quality Group Classifications and Standards for Water-,
Steam-, and Radioactive-Waste-Containinq Components of Nuclear Power Plants.
Seismic Design Classification
cQUESTION_l(a).
(Continued) During plant operation,Section XI of the ASME Boiler and Pressure Vessel Code, "Rules for Inservice Inspection of Nuclear Power Plant Components" provides guidance on inservice inspection of components and inservice testing of pumps and valves which are safetv-related; this is due to the fact th~t Surry was constructed prior to the development of ASME Section III which is applicable to Safety Related Systems today.
The construction codes and standards applicable to the auxiliary feedwater system and the safety-related portion of the main feedwater system at Surry Units 1 and 2 are as follows:
- 1.
Portions of main feedwater piping -
USAS B31.l.O - 1967 supplemented by ASME Code Case N-7.
Auxiliary feedwater piping -
USAS 831.1.0 - 1967.
- 2.
Pumps, such as auxiliary feedwater pumps - manufacturer's standards
- 3.
Valves - manufacturer's standards and USAS B31.l.O - 1967 and related standards; such as 816.5.
The construction codes and standards applicable to the nonsafety-related portions of the condensate and feedwater system at Surry Units 1 and 2 are as follows:
- 1.
Condensate and feedwater piping -
USAS B31.1.0 - 1967 Power Piping Code.
- 2.
Pressure vessels, such as feedwater heaters -
ASME Boiler and Pressure iessel Code,Section VII(, Pressure Vessels.
- 3.
Pumps, such as condensate and feedwater pumps, and steam turbines - manufacturer's standards.
- 4.
Valves - manufacturer's standards and USAS B31.1.0 - 1967 and related standards, such as B16.5.
(
Ii e
e U.S. NUCLEAR REGULA TORY COMMISSION NUREG-0800 (Formerly NUREG-75/087)
STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION 10.4.1 MAIN CONDENSERS REVIEW RESPONSIBILITIES Primary - Power Systems Branch (PSB)
Secondary - None I.
AREAS OF REVIEW The main condenser (MC) system is designed to condense and deaerate the exhaust steam from the main turbine and provide a heat sink for the turbine bypass system.
L
- 2.
The PSB reviews the performance requirements of the main *condenser for both direct and indirect cycle plants during all operating conditions.
Emphasis will be placed on the review of direct cycle facilities with regard to the prevention of loss of vacuum, corrosion and/or erosion, and hydrogen buildup.
The PSB reviews the design of the MC system with respect to the following:
- a.
The means to detect, control and facilitate correction of the leakage of cooling water into the condensate; to detect radioactive leakage into or out of the system; and to preclude accidental releases of radioactive materials to the environment in amounts in excess of established limits.
- b.
Instrumentation and control features that determine and verify that the MC is operating in a correct mode.
- c.
The means provided to deal with flooding from a complete failure of the MC and to preclude damage to safety-related equipment from the flooding.
- d.
The capability of the MC to withstand the blowdown effects of steam from the turbine bypass system.
In the review of the Main Condenser, the PSB will coordinate other branch evalua-tions that interface with the overall review of the system as follows.
The Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN Standard review plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsible for the review of applications to construct and operate nuclear power plants. These documents are made available to the public as part of the Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard *re....ew plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The standar.d review plan sections are keyed to the Standard Formm and Content of Safety Analysis Reports for Nuclear Power Plants.
Not all sections of the Standard Format have II corresponding review plan.
Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new informa-tion and experience.
Comments end suggestions for improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C. 20555.
V i.
r !.
Effluent Treatment Systems Branch evaluates the inventory of radioactive contami-nants in the MC during power operation and during ~hutdown as part of its primary
' review responsibility for SRP Section 11.5.
The Materials Engineering Branch, upon request of PSB, evaluates the adequacy of the materials of construction, the methods used to reduce the corrosion and/or erosion of MC tubes and compo-nents, the permissible cooling water inleakage, and the allowed time of operation with inleakage without affecting condensate/feedwater quality for safe reactor operation.
The Auxiliary Systems Branch determines that safety-related systems and structures are protected from the effects of flooding as part of its primary review responsibility for SRP Section 3.4.1.
The procedures and Test Review Branch determines the acceptability of the preoperational and startup tests as part of its primary review responsibility for SRP Section 14.0.
The reviews for fire protection, technical specifications, and quality assurance are coordinated and performed by the Chemical Engineering Branch, Licensing Guidance Branch, and Quality Assurance Branch as part of their primary review responsibility for SRP Sections 9.5.1, 16.0, and 17.0, respectively.
For those areas of review identified above as being part of the primary review responsibility of the other branches, the acceptance criteria necessary for the review and their methods of application are contained in the referenced SRP section of corresponding primary branches.
II.
ACCEPTANCE CRITERIA Acceptability of the design of the main condenser system, as described in the applicant 1s safety analysis report (SAR), is based on meeting the requirements of General Design Criterion 60 (GDC 60) and on the similarity of the design to that of plants previously reviewed and found acceptable.
The design of the Main Condenser System is acceptable if the integrated design of the system meets the requirements of GDC 60 as related to failures in the design of the system which do not result in excessive releases of radioactivity to the environment.
In addition, GDC 60 is satisfied if the system is designed such that failures do not cause unacceptable condensate quality, or flooding of areas housing safety-related equipment.
III.
REVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to determine that the design criteria and bases and the preliminary design meet the acceptance criteria given in subsection II.
For the review of operating license (OL) applications, the procedures are used to verify that the initial design criteria and bases have been appropriately implemented in the final design as set forth in the final safety analysis report.
The reviewer will select and emphasize material from this SRP section as may be appropriate for a particular case.
The primary reviewer will coordinate this review with other branches* areas of review as stated in subsection I.
The primary reviewer obtains and uses such input as required to assure that this review procedure is complete.
- 1.
The SAR is reviewed to determine that the system description delineates the main condenser system capabilities including the minimum system heat transfer and system flow requirements for normal plant and turbine bypass operation.
Measures provided to prevent loss of vacuum, corrosion and/or erosion of MC tubes and components, and hydrogen buildup in the MC are 10.4.1-2 Rev. 2 - April 1981
(
e reviewed, with particular emphasis on these measures for direct cycle
, (boiling water reactor) plants.
System performance requirements are reviewed to determine that they satisfactorily limit possible system degra-dation conditions (e.g., leakage, partial loss of vacuum) and describe the procedures that are followed to detect and correct these conditions.
The SAR is also reviewed to determine that any allowed MC system degraded operation does not have an adverse effect on the reactor primary system or secondary system in the case of pressurized water reactors.
- 2.
The reviewer evaluates the MC system design to verify that:
- a.
Means have been provided for detecting, controlling*and correcting condenser cooling water leakage into the condensate.
- b.
The permissible cooling water inleakage and time of operation with inleakage are provided to assure that condensate/feedwater quality can be maintained within safe limits.
- c.
Measures have been provided to detect *radioactive leakage into and out of the MC system and to preclude unacceptable accidental releases of radioactivity to the environment from the system.
- d.
The system is provided with instrumentation and control features that determine and verify that the MC is operating in a correct mode.
- 3.
The reviewer uses engineering judgment and the _results of failure modes and effects analyses to determine that:
- a.
The failure of a main condenser and the resulting flooding will not preclude operation of any essential systems.
Reference to sections of the SAR describing plant features and the general arrangement and layout drawings will be necessary, as well as the SAR tabulation of seismic design classifications for structures and systems.
Statements in the SAR that verify that the above conditions are met are acceptable.
- b.
The system, in conjunction with the main steam system, has provisions to detect loss of condenser vacuum and to effect isolation of the steam source.
For direct cycle plants, it will be acceptable if the detection system in the MC can actuate the main steam isolation valves to limit the quantity of steam lost from the condenser.
- c.
Design provisions have been incorporated into the MC that w1ll preclude component or tube failures due to steam blowdown from the turbine bypass system.
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient inform~tion has been provided and his review supports conclusions of the.following type, to be included in the staff's safety evaluation report:
The main condenser system (MC) includes all components and equipment from the turbine exhaust to the connections and*interfaces with the main condensate and other systems.
The scope of review of the main condenser system for the plant included layout drawings, piping and instrumentation diagrams, and descriptive information for the main condenser system and supporting systems that are essential to its operation.
10.4.1-3 Rev. 2 - July 1981
The basis for acceptance of the main condenser system in our review was conformance of the design, design criteria, and design bases to
'the Commission's regulation as set forth in GDC 60.
The staff cdncludes that the main condenser system design is acceptable and meets the requirements of GDC 60 with respect to failures in the design of the system which do not result in excessive releases of radioactivity to the environment.
The applicant has met this require-ment by providing radioactive monitors in the system to detect leakage into and out of the main condenser.
- V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
VI.
REFERENCES
- 1.
10 CFR Part 50, Appendix A, 11General Design Criterion 60, 11Control of Releases of Radioactive Materials to the Environment."
- 2.
Regulatory Guide 1.68, 11 Initial Test Programs for Water-Cooled Reactor Power Plants.
11 10.4.1-4 Rev. 2 - July 1981
,QUESrION_l(b).
ANSWER Are these requirements different than those applicable to other portions of the feedwater and steam lines that are closer to the steam generators and reactor vessel?
If so, whv 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 difference?
The construction codes and standards an~ regulatory requirements applied to the safety-related an~ nonsafety-related portions of the feedwater and steam svstems on PWR's are described in response to Question l(a)~ above.
Section XI of the Code currently does not contain a requirement to explicitly measure wall thickness to detect thinninq.
Weldments are tnspected by non-destructive examination~ to determine if indications are within allowable* limits.
The requirements noted above for the safety-related portions are appropriate since these lines (i.e., auxiliary feedwater) are relied upon to mitigate accidents resulting from failures in other lines.
In general, this distinction appears to still be appropriate.
However, the erosion-corrosion degradation and other failure mechanisms in nonsafety-related systems are under review to determine how on-site non-radiological injuries or fatalities which result from failures in such lines should be dealt with by the NRC.
If the results of such review indicate that NRC should play a more active role in protecting on-site personnel from non-radiological hazards, such systems could receive more attention in initial design and specific requirements to detect wall thinning and the inspection of weldment for flaws at critical locations could be added as part of an inservice inspection program.
. QUESTION_l(c).
ANSWER.
If a failure in the feedwater p1p1ng occurrPrl at a similar location, e.g., between the condenser and fePdwater pipinq ir a Boiling Water Reactor nuclear power plant, could radioactivity be released to the environment?
Yes.
If a feedwater pipe break occurred outside of containment, some fraction of the radioactivity in the water in the feedwater piping and main condenser would be released to the turbine building.
Isolation valves in the feedwater system would prevent backflow of water from the reactor vessel.
Since the turbine building is not designed as a containment structure, activity released to the turbine building is assumed to immediately enter the environment.
However, as discussed in the response to Question l(c)(i), no significant radiological cons~quences would be expected in the surrounding area.
- Q U E S T. I O N _ 1 ( c ) (.il.:.
ANSWER.
If so, how much rndioactivity could be released and what would be the consequences to the surrounding area?
Accident analyses demonstrate that no fuel damage would occur as a consequence of a feedwater pipe break outsirle of containment because other engineered safety systems would supply the core with coolant water.
While we don't have a specific calculation of the occupational exposure for a comparative pipe failure in the turbine building of a BWR, the amount of radioactivity available for release to the environment is limited to the radioactivity of the water in the condenser and the feedwater piping which is controlled by the plant Technical Specifications.
Tables 1 and 2 (attached) show the estimated radioactivity release and thyroid dose consequences for a typical boiling water reactor assuming that the activity released from a feedwater line break immediately enters the environment.
The calculated doses are for an individual at the site boundary are far below the guideline values of 10 CFR Part 100, and therefore no significant radiological consequences would be expected in the surrounding area.
TABLE 1 FEEDWATER LINE BREAK (REALISTIC ANALYSIS)
ACTIVITY RELEASE TO ENVIRONMENT, Ci Isotope Activity I-131 2.64E-2 I-132
- 1. 54E-l I-133 l.14E-1 I-134
- 1. 97E-1 I-135 l.14E-1 Total equivalent 5.35E-2 I-131 TABLE 2 FEEDWATER LINE BREAK RADIOLOGICAL EFFECTS Site boundary (2-hour dose)
Low population zone (30 day dose)
Thyroid, rem
- 1. 73E-3
<<1*
- Estimated based upon calculated 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose= l.32E-4 10 CFR Part 100 Guideline Values 300 rem to the thyroid 300 rem to the thyroid
e
,QUESTION_l(~)(ii).
ANSWER.
How are these areas of the feedwater and steam lines classified in Boiling Water Reactors?
Regulatory Guide (R.G.) 1.26, Standard Review Plan (SRP) Section 3.2.2, (Attached) and 10 CFR 50.55 provides the staff's criteria for classifying the main steam line and the feedwater line from the reactor up to and including the outermost isolation valve as Quality Group A (ASME Section III, Class 1).
R.G. 1.26 also classified the main steam line up to but not including the turbine stop valve and bypass valves as Quality riroup B (ASME Section III, Class 2) (See Table A.l, SRP 3.2.2).
Alternatively, for BWRs containing a shut-off valve (in addition to the two containment isolation valves) in the main steam line and in the main feedwater line, Quality Group 8 standards should be applied to those portions of the steam and feedwater systems extending from the outermost containment isolation valve up to and including the shutoff valve (See SRP 3.2.2).
Weldments in the steam and feedwater systems that are classified as Quality Groups A and Bare subject to periodic inservice inspection in accordance with Section XI of the ASME Code per 10 CFR 50.55(a)(g).
. QUES~ION_l(c)(iii).
ANSWER.
In view of the Surry accident, do you think that the classifications of these areas of the power plant (including the steam turbine, condenser and feedwater pumps) are appropriate?
The present classification of the steam and feedwater lines is appropriate in general.
Consideration will be given to incorporating the experience from Surry into the ASME Section XI inspection requirements for those portions of the systems classified as Quality Groups A and B.
Inspections of Quality Group D systems have not been reouired since failures of these systems are considered in the facility design and the consequences to public health and safety from a radiation exposure are well within 10 CFR Part 100 guidelines.
However, as discussed in the response to Questions l(b) and l(d), a determination will be made on whether nonsafety-related ~iping should also have inspection requirements.
, QUES:f ION_l ( d).
ANSWER.
What additional requirements could be applierl to the feedwater lines, steam lines, steam turhine, feedwater pumps, condenser and related equipment to improve the safety of nuclear plant operation?
Consideration will be aiven to periodically monitoring the pipe wall thickness in feedwater lines and other lines (both safety-related and nonsafety-related).
Although t~e concept is simple, such requirements would have to he specified with care to avoid testing literally miles of line.
Current inservice inspection of safety-related piping accordi.ng to Section XI of the ASME Boiler and Pressure Vessel Code is done in the vicinity of butt welds which have been sites for cracking in piping.
An analogous determination must be made for the case of wall thinning by an erosion-corrosion mechanism.
Irispection done to date* under programs either instituted prior to the Surry event or conducted in response to it, have revealed wall thinning of varying degrees in several piping systems in some plants.
That pattern must be evaluated in order to prepare a meaningful inspection program.
In addition, as discus.sed in the response to question l(b), a determination will be made on whether non-safety related piping should also be included in such programs.
The above addresses only the issues related to erosion and erosion-corrosion in safety-related piping systems carrying single phase fluids or two phase (steam water).
The broad question of what additional requirements could be applied to other equipment (nonsafety-related piping, steam turbine, feedwater pumps, condenser, etc.) whose failure does not have direct consequP.nces on the release of radioactive material must be reviewed furiher.
Design basis accident reviews, assuming failure of this equipment, indicate that the plants can reach safe shutdown in the event of such failures.
, Q U E ST I O N _ 1 ( e ).
ANSWER.
Does the Commission plan to make any changes in its regulatory requirements for Surry or other nuclear power plants in ord~r to implement less0ns learned from the Surry accident?
At this time, the issue of regulatory changes for equipment in this part of the plant is still under study.
The feedwater line which failed at Surry Unit 2 is not safety-related and is not covered by detailed NRC construction and periodic examination requirements.
We are reviewing the implications of accidents which have minimum or no radiological impact.
Based on the outcome of that review, the Commission may find it advisable to make changes in the regulatory requirements.
See also response to Questions l(b) and l(d) above.
- QUESTION_~.
ANSWER.
The NRC team report cited erosion-corrosion induced thinning of the pipe metal as the cause of the failure at the Surry Station.
Do the design, construction, maintenance and integrity monitoring codes, standards, or other regulations applied to nuclear power plants adequately provide for finding or make allnwances for deterioration of plant components and piping in service?
If not, does the Commission plan any regulatory changes to incorporate these factors in plant design, inspection and maintenance requirements?
The ASME Standards typically used for piping construction in nuclear plants are Section III of the Boiler and Pressure Vessel Code and Power Piping and ANSI/ASME B31.1, of the ASME Code for Pressure Piping, 831:
ANSI/ASME B31.1 does explicitly call for consideration of both erosion and corrosion in the desiqn process.
Neither of the above standards provides for requirements or offers explicit guidance on the important parameters to consider to avoid erosion-corrosion in initial designs.
Inspection of the nonsafety-related portion of the feedwater line during operation for either wall thinning or weldment flaws is not a requirement of B31.1.
Currently, there are no standards covering periodic monitoring for such piping.
Inspection requirements for safety-related piping are contained in Section XI of the ASME Boiler and Press*ure Vessel Code.
Section XI of the Code currently does not contain a requirement to explicitly measure wall thickness to detect thinning.
ASME has been formally requested by NRC to review its Codes and Standards applicable to both fossil and nuclear plants Tor appropriate changes to address the erosion and erosion-corrosion processes. both in the initial design process and later in integrity monitoring.
Regarding the plans by the NRC to make regulatory changes in design, inspection and maintenance requirements, please refer to the response to Question l(e).
Also, anv regulatory action by the NRC will consider actions undertaken by the nuclear industry and the national consensus standards.
- QUES"fION_3.
()UESTION_3(a).
ANSWER.
The two Surry Station nuclear units are very similar in design, nuclear reactor system and age.
The units also 11 share 11 some support and auxiliary functions.
In view of this depe~dency, can you explain why Unit 1 was not shut down immediately when the failure occurred in Unit 2?
Whose responsibility was it to decide whether or not to shut down immediately?
In your view, should Unit 1 have been shut down immediately?
The feedwater pipe rupture occurred at approximately ?:20 p.m. on December 9, 1986.
At the time both units had been operating at 100% power.
Following an automatic scram because of closure of a main steam trip valve, the rupture of an 18-inch feedwater pipe occurred.
After coping with the immediate actions necessary to mitigate the accident and account for and assist injured personnel, t.he licensee placed the l'nit 2 on a cool down ramp of 50°F per hour at 6:00 p.m. The unit achieved cold shutdown at 7:00 a.m. on December 10, 1986.
Inspection of the failed pipe took place on the morning of December 10 after scaffolding had been erected to reach the failed pipe sections.
It was at this time that it became apparent that accelerated pipe wall thinning had occurred, and that due to the similarity of design and configuration between Units 1 and 2, Unit 1 could have the same problem.
The root cause of the ~ailure appeared to be an erosion/corrosion mechanism that was not fully understood.
When the Augmented Inspection Team (AIT) arrived on site at 9:30 p.m. on December 9, considerations regarding shut down of Unit 1 were discussed with the licensee.
Immediately after the pipe rupture event and throughout the evening of December 9, the plant staff was engaged in recovering from the accident and in bringing Unit 2 to a cold. shutdown.
At that time, the licensee did not want to undertake the added burden of the shutdown of the other unit, and reasoned that maintaining Unit 1 in a stable operational mode was_the safer course of action.
In light of the fact that the root cause of failure had not been determined and Unit 1 was operating normally, the AIT, with Regional concurrence, agreed with the licensee 1 s conclusion.
Since single phase systems (containing water but no vapor) have historically not been susceptible to the types of failure mechanisms found in wet steam systems, it was thought that some unique flaw in the material might have caused the rupture.
The licensee had taken the prudent action to rope of~ the similar
,QUESTION_3(a).
(Continued) areQS in Unit 1 and had stationed security personnel to prohibit general entry into the area.
In addition, safety systems in Unit l were not affected by the accident and no dependency on Unit 2 systems existed for Unit 1 safe shutdown.
Shortly after the inspection of the Unit 2 ruptured piping, when it was determined that general thinning of the pipe wall could also have occurred in Unit 1, the licensee decided to shut down Unit 1 at 12:30 p.m. on December 10.
The unit was placed on a power ramp-down at 5:30 p.m. and achieved cold shutdown in the morning of December 11.
In summary, the NRC believes the actions taken by the licensee were prudent and actions of a more immediate nature were not warranted.
It is the licensee's responsibility to provide the basis for continued reactor operation.
The NRC reviews such bases.
If we disagree with the licensee's decision regarding continued operations, an order by the NRC to shut down would be appropriate.
QUES7I0N_3(b).
ANSWER.
Should the NRC issue any new guidan~e for such situations?
The Commission does not currentlv contemplate issuing new guidance which would change the basic responsibilities of the licensee or the NRC's role in event response as a result of the Surry incident.
For a lar9e variety of events the NRC is informed immediately.
The Incident Response Center at Headquarters and the appropriate Region is activated to monitor the licensee's actions during the event.
If appropriate, a team of experts is sent to the affected plant site.
The potential interactions between units at a site have to be considered on a case-by-case basis.
The actions taken by the licensee following the Surry Unit 2 event were appropriate and would not be a basis for developing additional guidance.
The operation of each unit is controlled by. the Technical Specifications applicable to a particular unit.
The Technical Specifications are developed with the intent that the plant should not be allowed to operate when it is considered to be in a potentially unsafe condition.
It is not'desirable to develop generic guidance for shutting down the plant as each situation is unique and it is difficult to foresee every sitl1ation in developing the generic guidance.
- QUEST.I ON_1_.
QUESTI0N_4(a).
ANSWER.
Changes in the control room ventilation system were being implemented while the plant was running 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 qas.
Are NRC regulations adequate for modifications being performed whilP plants are operating?
Were these rules being observed at the time of the accident?
The modifications being performed at Surry were to a nonsafety-related system not described in the Final Safety Analysis Report (FSAR).
Where modifications involve a change in the facility as described in the FSAR, 10 CFR 50.59 requires that an evaluation be performed by the licensee prior to making the modifications to determine that the modifications do not involve an unreviewed safety question or change in the technical specifications.
If the modification acttvities render a safety system incapable of performing its intended design safety function, the technical specifications require tha~ the system be declared inoperable.
Depending upon the particular safety system which has been declared inoperable, the technical specifications may require the plant to be shutdown within a specified time period.
Given this regulatory framework, NRC believes that, with proper implementation of the technical specifications and 10 CFR 50.59, the regulations are adequate with respect to modifications for safety systems.
The ventilation system changes that were being made at the time of the accident were not in a system described in the FSAR, nor did they directly impact the integrity of the control room envelope.
Consequently, these changes did not require an analysis under 10 CFR 50.59, nor were they a violation of the technical specifications.
Therefore, the licensee was in compliance with the regulations for the control room modifications which were being performed.
As discussed in the answer to Question 4(b) the operators were not following the procedures implicitly -- see answer to Question 4(b) for the details.
- QUESiI0N_4(b).
ANSWER.
Do you feel that different procedures should have been used?
Is the Commission considP.ring any regulatory changes to prevent ongoing modification work from compromising operational safety?
The accident resulted in the disabling of the key card security system for the control room doors.
The actions following the accident and the plant response to it required a considerable number of entries into the control room.
The licensee apparently believed the actions to block the doors open and post a guard were proper.
The doors remained open until the operators discovered that there was a release of carbon dioxide ~nd that the doors were allowing the carbon dioxide to reach the control room.
Using hindsight, the control room envelope could possibly have been better maintained by having the guards maintain the control room doors closed and opening them only to allow access as required for the accident mitigation actions.
The operators*
decision to close the control room door and establish the safety grade pressurization and ventilation systems, once the problem was discovered, was the proper action to restore habitability.
Tbe Commission is* not considering any regulatory changes to prevent ongoing modification work from compromising operational safety.
Compliance with plant technical specifications and 10 CFR 50.59 are considered adequate to ensure that required safety systems are maintained operable at all times~
,QLJES'l"10N_5.
QUESTI0N_5a.
ANSWER.
The NRC inspection team repor~s indicate the action was initiated by an improperly maintained valve.
Does it seem appropriate that the plant was allowed to operate with this valve not functioninq properly?
Are there adequate inspection requirements for such valves?
As a point of clarification, the NRC inspection team report did not describe the valvP. as being improperly maintained; it was improperly assembled during overhaul in November 1986.
After the overhaul thP. valve was tested in accordance with the Technical Specifications to verify its ability to close within the required time.
This valve is required to close in the event of a main steam line break in order to isolate the steam generator from the break location.
Even in its improperly assembled state, the valve would and did close properly as confirmed by testing in order to carry out this safety function.
The* problem was that because of its improper assembly it was susceptible to premature closure with normal steam flow in the pipe.
There are ci.dequate inspection requirements for such valvP.s.
The Technical Specifications require a closure test (valve must close within five seconds) and a test to assure that the valve disc is free to move.
These tests verify that the valve will be ahle to carry out its safety function of rapid closure to mitigate the effects of a steam line break.
Even in its improperly assembled state, the valve was capable of carrying out this function.
The deficiencies lie in the procedure for valve assembly, and the inspection required after assembly.
The procedure was inadequate in that it did not prevent, nor did the post-maintenance testing discover, the improper assembly of the valve.
In addition, the maintenance procedure used to overhaul the valve was not correctly followed and non-routine work associated with the overhaul not adequately documented.
Because of these deficiencies, the licensee was issued a Notice of Violation.
The deficiencies have been corrected.
yUES7I0N_5(b).
ANSWER.
Does the Commission plan any regulatory chanoes as a result of the maintenance deficiencies discovered during the investigation of this accident?
The Commission does not plan any changes as a result of the maintenance deficiencies discovered during the investigation of this accident.
The Commission recognizes the importance of maintenance in the safe operation of nuclear power plants and has focused added attention on these practices in the industry.
The Commission is considering directing the staff to develop a policy statement that would emphasize the Commission's concern with overall industry maintenance performance and to indicate that, to the eitent the self-improvement initiatives are effective and implemented on an industry-wide basis, the NRC would defer development of new maintenance re~uirements.