ML20058E574
| ML20058E574 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 07/28/1982 |
| From: | Paton W NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | SINCLAIR, M.P. |
| References | |
| ISSUANCES-OL, NUDOCS 8207300099 | |
| Download: ML20058E574 (64) | |
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07/ 28/82 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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CONSUMERS POWER COMPANY Docket Nos. 50-329 OL 50-330 OL (Midland Plant, Units 1 and 2)
NRC STAFF PARTIAL RESPONSES TO INTERROGATORIES SUBMITTED BY INTERVENOR SINCLAIR TO THE NRC STAFF ON JUNE 18, 1982 Before responding to interrogatories, the Staff wishes to clarify a matter concerning the numbering of contentions submitted by Intervenor Sinclair in October 1978. There is an obvious discrepancy in the i_
numbering of those contentions.
Following contention 45 there are contentions numbered 45 through 48. Then the numbering goes back to 44 and continues through 56. The Staff suggests that there will be less confusion if the contentions beginning with the second #45 (which appears on page 30 of the August 1978 contentions) be renumbered 46 and that all contentions from that point on be renumbered sequentially. Affidavits supporting the Staff's responses will be filed with a supplement to these responses.
The Staff provides no response at this time to Interrogatories 2a, 6, 8, 13a, b, c, f, 16a, b, c, d, f, 22, 23, 30, 36a and b.
Interrogatory 1 Provide the draft copy of NUREG-0410, "The NRC Program for the i
Resolution of General Issues Related to Nuclear Power Plants" (1978).
NRC Response to Interrogatory 1.:
l A copy of NUREG-0410 is provided.
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Interrogatory 2 Contention 6 deals wikh the paof quality control ' record of both the Applicant and the architect-engineer both at Palisades and Midland. As
~ ;ne. Board has requested, discovery, questions are to be directed to '
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,, current operation of the Quality Assurance. program (including the alleged
)' doctoring" of welding certificates) 7
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Provide all the documentation on the doctoring" of welding a.
certificates st Midland available to the NRC staff at this time.
b.
The rlur' rent status of the QA program has been most recently commented on pj the ACRS 10mtheir letter of June 11, 1982. Will the QA audit recommegdad by the ACRS be undertaken?
c.
What assurance does the public hsve that it will be an independent audit?
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d.
List the procedures by which the Quality Assurance issues that
, the-ACRS has raconsnended to be reviewed will be. undertaken.
Will there be opportunity for third partye' view of these r
e.
' procedures and the results?
NRC Response to. Interrogatory 2.b.:
j The Staff's reply to this ACRS request wapprovided by Section 19(1) of Supplement No. I to the SER.
More recently, with respect to this ACRS rp.%sst, the staff J
submitted a letter-to the appUcant, dated July 9,1982, requesting that
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'the appli' cant b'egin arJevaluation leading to an independent assessment of s Midland's, design adequacy.
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At this time, the staff awaits the applicant's, reply.
NRCzResponse to Interrogatory 2.c.:
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l-As' indicated in Response ?.b., the" staff c!vaits the applicant's
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l reply ~whether an independent design verification will he perfortned for j
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i the Midland Plant.
NRC Response to Interrogatory 2.'d.:
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. As can be seen by Response 2.b., the applicant's intent and the progress of implementation of this ACRS request has not proceeded to the stage that details of the procedures or review of results are known.
NRC Response to Interrogatory 2.e.:
As can be seen by Response 2.b., the applicant's intent and the progress of implementation of this ACRS request has not proceeded to the stage that details of the procedures or review of results are known.
Interrogatory 3 Contention 7 deals with the fact that the Applicant has distorted and suppressed the truth regarding important new information in proceed-ings before the Comission.
(Exhibits 24, 25, Suspension hearings, Dec. 1, 1976). The Applicant has continued to conceal important information, such as the failure to advise the NRC about the Administra-tion building settlement, the material false statement listed in the December 6,1979 Order that initiated the OM-OL proceedings, and four other false statements in Appendix A of the December 6 Order.
a.
In view of their history of concealing the truth, what assurance does the NRC staff have that further soils remedial work, as approved by the Construction Permit Amendment #3 and NRC's May 25, 1982 letter to Mr. Cook, will proceed with due regard for public health and safety?
Provide documentation.
b.
The Board's April 30, 1982 Memorandum and Order calling for the Amendment to the Construction Permifs expressed " doubt whether, in the absence of Staff review and approval, Consumers would carry out certain remedial soils activities using appropirate QA procedures and principles."
(p 14-15) What events happened between that April 30, 1982 Memorandum and Order arid the May 25, 1982 letter to Mr. Cook granting permission to proceed with Phase,II remedial work?
c.
Who were all the people involved in making this decision hN drawing up the May 25, 1982 letter to Mr. Cook? Provide documentation that substantiate the validity for coming to the decision that led to the May 25, 1982 letter rto Mr. Cook.
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d.
Did any members of the staff not agree with this letter? Who are they? Document their concerns.
e
. NRC Response to Interrogatory 3.a.:
The Staff objects. The premise of the interrogatory is not a fact of record and the interrogatory itself encompasses all issues before the Board.
NRC Response to Interrogatory 3.b.
Between April 30, 1982 and May 25, 1982, the following events related to the May 25, 1982 letter occurred:
1.
In accordance with the Board Order, the Staff defined an acceptable quality assurance plan and program for soils remedial work (enclosure 3 of the May 25, 1982 letter).
2.
The Staff provided (enclosure 7 of the May 25, 1982 letter) requirements for, and noted its acceptance of, Drawing 7220-C-45 for defining the areas around safety related structures and systems within which this Memorandum and Order would apply.
3.
Pursuant to the Board's Memorandum and Order of April 30, 1982, and a request by the applicant dated May 10, 1982, the Staff provided (enclosures 4 and 6 of the May 25, 1982 letter) clarification regarding its prior approvals.
4.
The NRC Staff completed documentation on its review of certain technical matters for acceptance of " phase 2" underpinning.
The phase-2 technical review matters and the conditions resulting from Staff review are identified by enclosure 2 of the Staff's May 25, 1982 letter.
l 5.
Staff review of technical and quality assurance aspects for installation of deep-seated benchmarks and relative-absolute instrumentation for monitoring the auxiliary building underpinning work was completed and accepted by enclosure 5 of the May 25, 1982 letter.
6.
The Staff also identified (enclosure 8 of the May 25, 1982 letter) the additional infomation needed to complete the review of remaining soils remedial work.
NRC Response to Interrogatory 3.c.
The May 25, 1982 letter documents staff approval of phase 2 underpinning activities, including the basis and conditions thereto. The
. staff decision is consistent with the Board's Memorandum and Order of April 30, 1982. The people involved in making this decision and in drawing up the May 25, 1982 letter are:
Joseph Kane Lyman Heller George Lear Frank Rinaldi Franz Schauer Raymond Gonzales James P. Knight Richard Vollmer Darl Hood Elinor Adensam Robert Tedesco Darrell G. Eisenhut John Gilray Ross Landsman Charles Norelius Walter Haass Cordell Williams In arriving at its decision, the Staff had the benefit of advice of several technical consultants and legal staff.
NRC Response to Interrogatory 3.d.:
The NRC Staff is unaware of any staff disagreement with the May 25, 1982 letter.
Interrogatory 4 Contentions 20 and 21 on the nuclear fuel cycle and the lack of a method to store nuclear waste should now be admitted for discovery since the U.S. District Court of Appeals struck down as invalid the S.3 Table (April 27,1982) on which the NRC was relying for compliance with NEPA.
I am resubmitting these issues in my amended list of contentions.
Response
No response required.
Interrogatory 5 Contention 24 is now the basis for the on-going soil settlement hearings.
1
- Response No response required.
Interrogatory 6 Contention 27 deals with the lack of an adequate emergency evacuation plan at Midland.
a.
Who will decide when an emergency evacuation is necessary?
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b.
The area warning system has frequently malfunctioned. How will people be convinced it is a real emergency?
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c.
How high w uld radiation doses have to be before evacuation of a 10 mile zone is ordered?
d.
Will the radiation dose limits for evacuation vary for men, l
l women, children, pregnant women and infants? In what way?
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e.
When Dow had a major chlorine leak several years ago, all the i
communications to the plant were Jamed with people calling in or trying to call out. How will this be avoided if the Midland nuclear plant has an emergency?
f.
Have parents been consulted about how their children should be taken care of if they are in school during an emergency?
g.
During the Dow chlorine leak, people were driving into the cloud to look for their sick relatives, children and pets. Do the emergency plans allow for known human reaction patterns shown in past emergency situations? What are these plans? Provide documentation.
1 h.
How much time does the safe shutdown of all critical processes at Dow require?
1.
What protection will the, workers have who must stay during an emergency to complete the shutdown process?
- j. What special training for an emergency will these workers have?
k.
What guarantees do you have that they will stay as long as needed during an emergency?
'4.-
c 1.
Will uncontamiriated food and water supplies be kept available for their use for several days in the event of an emergency?
m.
What are the host cities to which people in a 10 mile radius will be evacuated? Have they been notified and prepared for this?
. n.
How will people who do not understand the English language be notified?
o.
How will people in nursing homes be evacuated?
p.
How will people in hospitals be evacuated?
q.
How will people who do not own cars be evacuated?
r.
Hcw many beds for treatment of radiation poisoning does the Midland Hospital have? Bay City General? Saginaw General?
s.
What plans are in place to deal with changes in wind direction after evacuation has begun? How will people be notified of this change?
t.
Are there segments of the population for whom no evacuation plans can be made? Who are they? Why can't they be evacuated?
u.
The NRC says there could be radioactive fellout as far as a 50 mile radius. What protection will there be for residents beyond a 10 mile radius? Have their officials been included in the emergency planning process?
v.
Homeowners insurance policies specifically exclude coverage for loss due to a nuclear accident. Will homeowners be able to recover their losses from some other source, since the area could be uninhabitable for decades?
Response
Interrogatory 7 l
Contention 28 deals with the water hammer problem of pressurized water reactors of the Midland type. This problem is identified as one of the unresolved safety issues applicable to Midland 1 & 2 in the SER, C-4.
a.
Since other reactors are now operating without having this problem resolved, would failure to have this problem resolved be l
sufficient reason not to approve an operating license for Midland 1 and 27 b.
Given the same premise, would you allow the plants to operate at full power with this defect?
c.
What is the series of events in the reactors that will take place when and if the water hammer problem manifests itself?
. d.
What non-safety related systems can affect or initiate the water hammer problem? Provide documents that explain this interaction between the water hamer problem and non-safety related systems.
e.
Provide the most recent summary documents of the Task Force A-1 that indicate methods for resolving the water hamer problems.
f.
How will this unresolved safety problem affect the total power output of these nuclear plants?
g.
Has there been any incident in an operating reactor which raised this as a concern? Describe it and provide documents on the incident or incidents.
h.
Why is this an unresolved safe +.y problem?
NRC Response to Interrogatory 7.a.:
No. The staff's answer and basis for licensing and full power operation of the Midland Units 1 and 2 prior to resolution of this Unresolved Safety Issue (USI) are provided on page C-7 to the SER.
In general, for those USI's that are unresolved, the Staff has found that current NRC requirements related to the issue as found in the Regulations, the Standard Review Plan, and Regulatory Guides, supplemented with additional interim requirements on some issues, provide the basis for safe operation of the plants prior to the ultimate resolution of the unresolved safety issue. The staff's discussion in Appendix C of the SER of this and other unresolved safety issues that are generically unresolved references those sections in the body of the SER where a plant-specific assessment is conducted in accordance with the appropriate Regulation, Standard Review Plan or Regulatory Guide.
In general, failure to have a USI resolved generically would not be a basis for denying an operating license on a specific plant.
NRC Response to Interrogatory 7.b.:
a
. Yes. The staff's basis for allowing full power operation at the Midland units prior to resolution of a USI is discussed on page C-7 of the SER, as noted in response to Interrogatory 7.a. above.
NRC Response to Interrogatory 7.c.
Recent events at operating B&W plants with an internal auxiliary feedwater (AFW) feed ring of the same design as that at Midland have shown 3 marked susceptibility to internal damage of the feed ring as a result of water hammer.
From this, reduced cooling in the steam generators could occur as a result of inadequate AFW flow following a loss of normal feedwater flow.
No reduction in AFW flow or cooling of the RCS was actually observed at any of the operating B&W plants.
NRC Response to Interrogatory 7.d.
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A loss of normal feedwater flow due to failure of the main feedwater system for any reason will result in AFW initiation with a resultant ootential for water haniner in B&W steam generators with an internal AFW feed ring. Refer to Sections 7.3, 7.4, 10.4.9, and 15.2 of the Midland SER (NUREG-0793) for an evaluation and analysis of AFW initiation following a loss of nonnal feedwat flow.
NRC Response to Interrogatory 7.e.:
A copy of the staff's Task Agtion Plan A-1 to resolve this USI is provided.
NRC Response to Interrogatory 7.f.:
The staff's basis for allowing full power operation of the Midland units is discussed on psge C-7 of the SER, as noted in response to Interrogatory 7.a. above.
l NRC Response to Interrogatory 7.g.
Yes. A recent inspection of the Davis-Besse steam generators showed l
significant damage to the AFW waterbox and distribution ring inside the l
steam generator. As a result, Rancho Seco and Oconee Unit 3 perfonned inspections and found similar damage. As with Midland, the design of all three plants will be changed to eliminate use of the internal distribution ring and waterbox, and an external distribution ring of the same type as provided on operating B&W plants that have not had any known water-hamer problems will be added. Licensee Event Report (LER)82-019 dated April 30, 1982 plus supplemental infonnation attached to the LER provides a description of the incident for Davis-Besse. Revision 1 to o
LER 82-019 dated May 24, 1982 provides further details of the damage incurred at the Davis-Besse plant. Our meeting summary dated May 25, 1982 describes the planned repairs at Davis-Besse. LER 82-10 dated April 20, 1982 plus followups dated April 29, 1982, May 19, 1982 and May 20, 1982 describe the incident and proposed modifications for the Rancho Seco plant. For Oconee Unit 3, the incident and proposed repairs are described in a preliminary notification dated April 30, 1982 plus Reportable Occurrence Report R0-287/82-06 dated May 14, 1982 and its revision dated June 14, 1982. These LER's are provided.
NRC Response to Interrogatory 7.h.:
The staff's basis for classifying this issue as an Unresolved Safety Issue is provided in NUREG-0510, " Identification of Unresolved Safety Issues Relating to Nuclear Power Plants," Pages 9 through 18. A copy of NUREG-0510 is provided.
Interrogatory 8
. Contention 29 deals with the failure of the design for the reactors to consider the effect of an asymmetric loading on the reactor vessel supports resulting from a postulated reactor coolant pipe rupture at specific locations.
a.
What is the precise way in which you have addressed this problem to meet the special design at Midland?
b.
Provide names and reports of contractors, consultants and documents of staff work for resolving this problem.
Response
Interrogatory 9 Contention 30 deals with the degradation of steam tube integrity.
Babcock and Wilcox (B&W) steam generator tube integrity is listed as one of the unresolved safety problems at Midland 1 & 2.
(SER,C-4) a.
Since other reactors are now operating without having this problem resolved, would failure to have this problem resolved be sufficient reason not to approve an operating license for Midland 1 & 27 b.
Given the same premise, would you allow the plants to operate at full power with this defect?
c.
What is the series of events in the reactor that will take place when and if the steam generator tube degradation problem manifests itself?
I l
d.
What non-safety related systems can affect or initiate the steam j
generator tube degradation problem? Provide documents that explain this interaction between the steam generator tube degradation problem and non-safety related systems.
e.
Provide the most recent summary documents of the Task Force A-3, A-4, and A-5 that indicate possible methods for resolving the steam generator tube degradation problem.
f.
How will this unresolved safety problem affect the total power output of these nuclear plants?
l g.
Has there been any incident in an operating reactor which raised i
this as a concern? Describe it and provide documents on the incident or incidents.
h.
Why is this an unresolved safety problem?
l l
. 1.
Provide documentation on corrosion problems at other operating B&W plants.
- j. Provide documentation to show how the type of corrosion that has occurred at the TMI-1 reactor steam generator while standing idle cannot occur at Midland.
NRC Response to Interrogatory 9.a.:
No. The staff's answer and basis for licensing and full power operation of the Midland units prior to resolution of this issue are provided on pages C-7 through C-9 of the SER, and in response to Interrogatory 7.a. above.
NRC Response to Interrogatory 9.b.:
Yes. See response to Interrogatory 9.a. above.
NRC Response to Interrogatory 9.c.:
Steam generator tube degradation would be detected by inservice inspection. The NRC requirements for inservice inspection of B&W plants are specified in standard tec.5nical specifications. Steam generator tubes are required to be plugged when the extent of the tube degradation extends to 40% of the total wall thickness.
During operation, steam generator tube degradation to the extent h
that excessive primary to secondary leakage occurs would be indicated to the operator by one or,more of the following symptoms:
a.
Decreasing reactor coolant pressure b.
Decrease in pressurizer level c.
Decreasing make-up tank level
<t d.
IncreasingMake-up flow e.
High steam line radiation level d
. f.
High condenser radiation level g.
Secondary coolant sample analysis h.
Increasing steam generator water level NRC Response to Interrogatory 9.d.
Any non-safety related system that can adversely affect secondary system water chemistry can increase the rate of tube degradation. At the Midland plant these include the circulating water system and the condensate storage system. Tube leaks in the main condenser could result in circulating water being introduced into the secondary water system which will result in improper chemistry control and may lead to faster tube degradation.
In the same manner failures to properly control the l"
condensate storage and transfer system chemistry which is used for secondary system makeup could result in improper steam generator water chemistry that may lead to faster tube degradation. A history and description of steam generator tube experience is provided in response to Interrogatory 9.g.
However, we are not aware of any documents that explain the interaction between the steam generator tube degradation problem and non-safety-related systems.
NRC Response to Interrogatory 9.e.:
1 A copy of the staff's A-3, 4, and 5 Task Action Plan to resolve this USI is provided. However, only A-5 is applicable to B&W steam generators.
NRC Response to Interrogatory 9.f.:
See response to Interrogatory 9.a. above.
NRC Response to Interrogatory 9.g.
. Incidents associated with degradation of steam generator tube integrity are identified and discussed in NUREG-0886 " Steam Generator Tube Experience" and NUREG-0916 " Safety Evaluation Report Related to the Restart of R-E. Ginna Nuclear Power Plant," (Docket No. 50-244). Copies are provided.
NRC Response to Interrogatory 9.h.:
See response to Interrogatory 7.h. above.
NRC Response to Interrogatory 9.1.:
Pertinent documentation on corrosion problems at operating B&W plants are contained in NUREG-0571 entitled, "Sumary of Tube Integrity Operating Experience With Once-Through Steam Generators," and NUREG-0886 entitled, " Steam Generator Tube Experience." These are provided.
NRC Response to Interrogatory 9.J.:
The basis for the conclusion that the type of corrosion that occurred in the TMI-1 steam generators while standing idle can not occur at Midland is that sodium thiosulfate will not be used in the containment spray system.
Interrogatory 10 Contention 31 deals with anticipated transients without scram (ATWS).
a.
Indicate precise ways in which this problem will be handled for Midland given the unique design of this plant and its interrelationship i
with The Dow Chemical Co.
l l
b.
Provide draft copy of NUREG-0460, Vol. 4, and documents of staff, consultants and contractors dealing with the resolution of this problem.
c.
Indicate all non-safety related systems that can affect ATWS.
. NRC Response to Interrogatory 10.a.:
The provisions of the Midland plants in dealing with ATWS events are discussed in section 15.6 of the SER. The process steam system by which steam is supplied to the Dow Chemical Company will not affect the likelihood or severity of postulated ATWS events.
NRC Response to Interrogatory 10.b.:
A copy of NUREG-0460, Vol. 4, is provided.
NRC Response to Interrogatory 10.c.:
ATWS events would be produced by the combination of an anticipated transient coupled with the failure of the reactor to scram. The scram systems are safety related, however, anticipated transients could be produced by control failures in non-safety related systems. These systems include the feedwater system, main steam system, makeup and purification system, non-vital electrical power systems and the intearated control system.
Interrogatory 11 Contention 32 deals with the questions of suitable safety margins for materials used for reactor vessel fabrication. Reactor Vessel Materials Toughness is listed as og of the unresolved safety problems in the SER, of C-4.
The questions of teactor embrittlement and the consequences of thermal shock have had increased attention by the NRC.
l a.
Provide all the documents and papers of Dimitri Basedekas who l
has stated that cracking of reactors will occur as a result of the l
embrittlement problem.
i b.
Provide documentation on the materials that went into the construction of the Midland reactors as well as the dates they were built and the dates they,were installed.
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e.
c.
Provide reasons and documentation why Unit I which was known to be among 9 defective reactors in the country with a high copper content in a major weld was permitted to be installed even though this defect was known for some time.
. d.
Provide evaluations of pressure vessel integrity at other reactors and how these compare with Midland.
e.
Provide any analysis of rapid cool downs and how they compare with Midland.
f.
What surveillance requirements are required for pressure vessels of the B&W Midland type?
g.
Provide documentation to show that these surveillance require-ments are adequate.
h.
Has there been any incident in any operating plant which raised a concern on this problem? Describe it and provide documentation.
1.
Provide documents to show why this is an unresolved safety problem.
NRC Response to Interrogatory 11.a.:
A copy of Mr. Basdekas's April 29, 1982, memorandum to the NRC
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Commissioners on this topic is provided.
NRC Response to Interrogatory 11.b.:
Materials that went into the Midland Units 1 and 2 reactor vessels are documented in Section 5.2 and 5.3 of the Midland Final Safety AnalysisReport(FSAR). Fabrication of the reactor vessels for Units 1 and 2 was completed at the Babcock and Wilcox factory in 1974 and 1975, respectively.
Both vessels were installed at the Midland site in 1978.
NRC Response to Interrogatory 11.c.:
At this time, the NRC staff does not believe the existence of a high copper weld in Midland Unit I reactor vessel beltline requires immediate action by Consumers Power Company.
Corrective action to ensure that the l
Midland Unit I reactor vessel maintains adequate safety margins depends on the degree of material embrittlement caused by neutron irradiation.
The NRC staff will be able to determine the degree of neutron l
embrittlement of the Unit 1 beltline reactor vessel materials through the l
. Midland Surveillance Programs (Midland FSAR Section 5.3) and the Babcock
& Wilcox Owners Group Surveillance Program (Babcock & Wilcox Report BAW 1543).
If the results of these programs indicate that reactor vessel safety margins are unsuitable during postula+.ed accident conditions, such as thermal shock event, the applicant has indicated that other actions, such as performing an inplace thennal anneal of the high copper weld in Unit I reactor vessel beltline, which is described in FSAR Section 5.3.3.8, could be performed.
Based on the Midland and Babcock & Wilcox Owner Group surveillance programs and the alternative actions which can be implemented to improve reactor vessel material properties, the NRC staff concludes that concern for reactor vessel embrittlement during postulated accident conditions have been adequately addressed.
NRC Response to Interrogatory 11.d.:
Pressure vessel integrity is discussed in the NRC staff's Safety Evaluation Report Section 5.3.3 for all applicants. Reactor vessel integrity is ensured by requiring all licensees to operate their reactor vessels in accordance with pressure temperature limits which are calculated based on the safety margins in Appendix G, 10 CFR Part 50.
The Midland pressure temperature limits will be based on the safety margins in Appendix G,10 CFR Part 50; therefore Midland's reactor vessel integrity will be equivalent to other nuclear facilities.
NRC Response to Interrogatory 11.e.:
The NRC staff has performed an analysis of the rapid cooldown which occurred at the Rancho Seco Nuclear Generating Station on March 20, 1978.
This analysis is documented in a memorandum from D. Eisenhut to V. Stello Jr. dated March 30, 1978.
a
. The limiting material in the Rancho Seco reactor vessel was fabricated using the same weld wire and flux as the limiting material in the Midland Unit I reactor vessel beltline and has equivalent chemical composition and fracture toughness properties. This indicates that the staff's conclusions concerning the Rancho Seco reactor vessel beltline materials are applicable to the Midland Unit I reactor vessel beltline materials.
NRC Response to Interrogatory 11.f.:
Surveillance requirements for all nuclear reactor vessels of the B&W Midland type are defined in Appendix H, 10 CFR Part 50.
NRC Response to Interrogatory 11.g.:
The staff's review of the Midland reactor vessel beltline surveillance program is documented in Section 5.3.1 of the Midland Safety Evaluation Report.
NDC Response to Interrogatory 11.h.:
The rapid cooldown incident at Rancho Seco on March 20, 1978 raised the staff's concern for reactor vessel integrity. This event is described in the staff's safety ev4)uation of the incident which is documented in a memorandum from D. Eisenhut to V. Stello Jr. dated March 30, 1978.
NRC Response to Interrogatory 11."i.:
See response to Interrogatory 7.h. above. However, the staff further notes that akpects of A-11 which dealt with pressurized thermal shock have recently been defined as a new USI, A-49. Documentation identifiying this as a USI is provided in SECY 81-687. A copy of
. SECY-81-687 dated December 8,1981, " Designation of Pressurized Thermal Shock as an Unresolved Safety Issue." is provided.
Interrogatory 12 Contention 33 addresses the necessity of reassessing the fracture toughness of the steam generator and reactor coolant pump support structure because of the potential for lamellar tearing and low fracture toughness of these materials. This has been identified as an unresolved safety problem in the SER, C-4.
a.
Since other reactors are now operating without having this problem resolved, would failure to have this problem rc. solved be sufficient reason not to approve an operating license for Midland 1 & 27 b.
Given the same premise, would you allow the plants to operate at full power with this defect?
c.
What is the series of events in the reactors that will take 5
place when and if the low fracture toughness and lamellar tearing problem manifests itself?
d.
What non-safety related systems can affect or initiate this problem? Provide documents that explain this interaction between this problem and non-safety related systems, e.
Provide the most recent summary documents of the Task Force A-12 that indicate methods for trying to resolve this problem.
f.
How will this unresolved safety problem affect the total power output of these nuclear plants?
g.
Has there been any incident in an operating reactor which raised this as a concern? Describe it and provide documents on the incident or incidents.
h.
Why is this an unresolved safety problem?
NRC Response to Interrogatory 12.a.:
No. The Staff's answer and basis for licensing and full power operation of the Midland units prior to resolution of this issue are provided on pages C-10 through C-11 of the SER, and in response to Interrogatory 7.a. above.
. NRC Response to Interrogatory 12.b.:
Yes. See response to Interrogatory 12.a. above.
NRC Response to Interrogatory 12.c.:
As discussed on pages C-10 and C-11 of the SER, the NRC staff does not believe lamellar tearing of reactor coolant pump or steam generator supports will occur at Midland. The staff has not evaluated the series of events in the reactors which would take place if it occurred.
NRC Response to Interrogatory 12.d.
The NRC staff is not aware of any non-safety systems which could affect or initiate lamellar tearing of reactor coolant pump or steam generator supports.
NRC Response to Interrogatory 12.e.:
A copy of the staff's Task Action Plan A-12 to resolve this USI is provided.
NRC Response to Interrogatory 12.f.:
See response to Interrogatory 12.a. above.
NRC Response to Interrogatory 12.g.
To the best of our knowledge, there have been no incident in operating reactors of failure of steam generator and pump supports related to lamellar tearing or low fracture toughness.
NRC Response to Interrogatory 12.h.:
See response to Interrogatory 7.h. above.
Interrogatory 13 Contention 34 deals with the actual and potential of snubber malfunction.
- a.
Provide documents on the methodology irnployed to determine the necessity for using snubbers as component supports in the Midland project.
b.
How does the snubber problem specifically apply to Midland?
c.
List the specific measures that will be taken to resolve this issue.
d.
What non-safety related systems can affect or initiate the malfunction of snubbers? Provide documents that explain this interaction between the snubbers that malfunction and non-safety related systems.
f.
How will this unresolved safety problem affect the total power output of these nuclear plants?
g.
Has there been any incident in an operating reactor which raised this as a concern? Describe it and provide documents on the incident or incidents.
h.
Why is this an unresolved safety problem?
NRC Response to Interrogatory 13.d.
We do not know of any non-safety-related systems that could affect or initiate the malfunctions of snubbers.
NQC Response to Interrogatory 13.g.
Snubbers are used in reactor plants to prevent safety-related components and systems from being overstressed during seismic events.
In 1974, hydraulic snubbers experipced leakage problems and were put under a surveillance program; but the same requirements were not applicable to mechanical snubbers.
In 1980, Hatch Unit No. 2 discovered that 42 of the 63 mechanical snub'bers by International Nuclear-Safeguard Corp. that were installed in the plant had locked up permanently. By lettersdatedNovembbr 20, 1980, and April 17, 1981. NRC has requested all operating reactors to include mechanical snubbers in the surveillance program by revising their Technical Specifications.
I&E also issued Bulletin 81-01.
. Four LERs describing recent incidents are provided.
NRC Response to Interrogatory 13.h This is not an unresolved safety problem.
Interrogatory 14 Contention 35 deals with pressure vessel integrity and the "significant uncertainties" in the ability to detect and adequately size flaws to assure continued integrity of the reactor coolant pressure boundary and to assess margin against failure under various plant conditions for the full life of the plant.
a.
What is the precise way in which the staff has addressed this problem for the Midland design?
b.
In the staff's opinion, has this been resolved for Midland?
Provide documentation.
c.
Since the accident at THI-2, it is known that failure probability of a reactor pressure vessel must be considered as a design basis accident. What is the course of events that will occur that can lead to such an accident?
d.
What is the precise probability for such an event for Midland 1
& 27 e.
Provide all documents on the ability to detect and adequately size flaws in the pressure vessel.
f.
Provide names of contractors, consultants and staff members who are responsible for this PRA.
g.
Identify any staff members or consultants who disagree with these views. Provide documents on their views.
h.
Has there been any incident in any operating plant which raised this issue as a concern? Describe it.
Provide documents to show how it was resolved.
NRC Response to Interrogatories 14.a. and 14.b.:
The detection and sizing of ultrasonic testing (UT) flaw indications during preservice and inservice inspections are being addressed on a generic basis as an integral part of several generic tasks.
5
. In the case of Midland, the detection and sizing of UT flaw indications is being addressed by requiring compliance with 10 CFR Part50,Section50.55a(g). The regulation requires compliance with the applicable edition and addenda of Section XI of the ASME Code and augmented inservice inspection programs for systems and components for which the Comission deems that added assurance of structural reliability is necessary. Since the regulation requires updating of inservice inspection programs throughout the service life of a plant, the regulation has provisions for granting relief from impractical examination requirements.
NRC Response to Interrogatories 14.c. and 14.d.
There were no aspects of the TMI-2 incident that required considering reactor vessel failure to be a design basis accident. The NRC position remains that reactor vessel failure should not be considered to be a design basis accident. There are many technical reasons for this position, even considering uncertainties in flaw detection.
Nevertheless, the probability of detecting flaws significant to the safety of the Midland vessel will be enhanced by the use of improved l
inspection methods for inservice inspection that are in accordance with the staff positions of Regulatory Guide 1.150 "Ultransonic Inspection of Reactor Vessel Welds During Preservice and Inservice Examinations."
Although the preservice inspections of the Midland vessels were perfomed before the Regulatory Guide was issued, and therefore was not explicitly followed, the methods used, in conjunction with the original fabrication inspections performed, are adequate to ensure freedom from original material and fabrication flaws. The use of improved methods covered by
. Reg. Guide 1.150 for inservice inspections will. significantly improve the probability of detecting any deleterious flaws developing during service.
NRC Response to Interrogatory 14.e.:
A pertinent document on the subject is Regulatory Guide 1.150 entitled " Ultrasonic Testing of Reactor Vessel Welds During Preservice And Inservice Examinations." A copy is provided.
NRC Response to Interrogatory 14.f.
Ne rkA has been performed.
NRC Response to Interrogatory 14.g.:
l l
We do not know of staff or consultant dissenting opinions regarding the ability to detect and adequately size flaws in the preservice
~-
inspection of vessels reported in SER Section 5.2.4.
NRC Response to Interrogatory 14.h.:
We do not know of any incident in any operating plant of a failure l
nf the reactor pressure vessel that was a result of the inability to detect and adequately size flaws.
l Interrogatory 15
)
i Contention 36 discusses the lack of systematic process to review different nuclear powe,r plant systems to determine their safety-related impact on other parts of the plant. Systems interactions is identified as an unresolved safety problem applicable to Midland 1 & 2 in the SER, C-4.
i a.
Since other reactors are now operating without having this problem resolved, would failure to have this problem resolved be l
sufficient reason.not to approve an operating license for Midland 1 & 2?
g b.
Given i.he same premise, would you allow the plants to operate at full power with this defect?
c.
Provide the most recent summary documents of the Task Force A-17 that indicate methods for resolving this problem.
1 1
. d.
How will this unresolved safety problem affect the total power output of these nuclear plants?
e.
Describe and document the incidents in operating B&W reactors where systems interaction was a concern.
f.
Why is this an unresolved safety problem?
NRC Response to Interrogatory 15.a.:
No. The staff's answer and basis for licensing and full power operation of the Midland units prior to resolution of this issue are provided on pages C-11 through C-13 of the SER, and in response to interrogatory 7.a.
NRC Response to Interrogatory 15.b.:
Yes. See response to Interrogatory 15.a. above.
NRC Response to Interrogatory 15.c.:
A copy of the staff's Task Action Plan A-17 to resolve this USI is provided.
NRC Response to Interrogatory 15.d.:
See Response to Interrogatory 15.a. above.
NRC Response to Interrogatory 15.e.
There are three events at operating B&W reactors that have selectively been of interest to the systems interaction program.
Fi rst, the TMI-2 event on March 28, 1979, is of interest both generally because of its consequences and specifically because of several humanly coupled interactions that contributed to the event. The event is well documented in the M. Rogovin and G. Frampton report on Three Mile Island, January 1980. Second, the Crystal River-3 event on February 26, 1980, is of interest because of the systems interaction where the Integrated Control system input, the PORV positioning, the instruments used for
. manual control of ECCS, and the entire Non-nuclear Instrumentation (NNI) power supply depended upon one + 24VDC line within the NNI power supply system. The event is documented in NUREG-0667, " Transient Response to Babcock and Wilcox Designed Reactors," May 1980. Thirdly, the Davis-Besse-1 event on April 19, 1980, is of interest because maintenance activities allowed an elimination of redundant power supplies that were supporting the decay heat removal function. Concurrent construction activities caused the loss of the working power supply and subsequently decay heat removal was lost for over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This third event is doci?-
mented in USNRC IE Information Notice 80-20, " Loss of Decay Heat Removal Capacity at Davis-Besse 1 while in a Refueling Mode," May 8, 1980.
NRC Response to Interrogatory 15.f.:
See Response to Interrogatory 7.h. above.
Interrogatory 16 Contention 37 deals with the absence of adequate design criteria for postulation of pipe breaks and protection therefrom.
a.
Precisely how does this lack of design criteria for pipe breaks apply to the Midland plant design?
b.
Provide names and reports of all contractors and consultants who have worked on this problem.
Provide summary documents on their work, c.
Provide documents of staff that worked on this problem, d.
Have any staff members or consultants disagreed with the criteria being used? Who are they? Describe the substance of their disagreement and provide documents on this, e.
What non-safety related systems can initiate or aggravate a pipe break problem?
f.
How will this interaction be monitored or controlled?
O.
NRC Response to Interrogatory 16.e.
Any non-safety related system itself can initiate a pipe break but cannot cause a pipe break in any safety related systems or any system necessary to mitigate the consequences of the initial pipe break.
Midland relies on non-safety grade isolation valves in the main steam system to mitigate the consequences of a pipe break in a seismic Category I steam line upstream of a main steam isolation valve (MSIV) coincident with a failure to close safety related MSIV in the steam line from the unaffected steam generator. This is acceptable, as discussed in Section 10.3 of the SER.
Interrogatory 17 Contention 38 deals with the inadequate analyses of main steamline break and the concerns regarding the capability of the equipment to survive such a break inside the containment, a.
Has a PRA been made for this problem? What is it? Provide documents.
b.
Precisely how does this apply to the Midland 1 & 2 design?
c.
Provide names of staff, contractors and consultants who have worked on this problem and their fpal reports for resolving this issue.
d.
Has any staff member, contractor or consultant disagreed with your final resolution of this; issue?
e.
If so, what were the reasons for their dissent? Provide documents on their reasons.
NRC Response to Interrogatory 17.a.:
As noted in Seckion 19(7) of Supplement 1 of the SER, the Applicant
's:-
e.
will provide a PRA of Midland,to the Staff in early 1983.
The staff is not aware of any analyses for the effects of main steamline breaks inside containment on the capability of equipment.
. NRC Response to Interrogatory 17.b.
The applicant has identified the worst-case MSLB with respect to 2
containment temperature response, to be a 12.22-ft break at 102 percent of full power; the applicant calculated a peak containment temperature of 458'F. The applicant performed the containment analysis for the postulated MSLB assuming two active failures, namely, the failure of one emergency diesel generator to start (mimimum containment heat removal capability) following a loss of offsite power and the failure of one main steam isolation valve to close. Assuming two active failures goes beyond staff requirements of assuming a single active failure and results in an l
l overly conservative prediction of the containment temperature response.
There is a further conservatism in the applicant's analysis in that the mass and energy release data were based on the assumption of no loss of offsite power (i.e., the reactor coolant pumps were assumed to remain operable to maximize the energy transfer from the primary to the secondary system).
The staff performed a confirmatory analysis of the containment temperature response for this MSLB, and the results confirm the acceptability of the applicant's analysis.
The staff's review also included a review of component thermal I
analyses. The applicant has not yet provided the component thermal analyses to justify the proposed qualification of safety-related components at temperaturet lower than the peak calculated containment atmosphere temperature. Staff evaluation of the capability of equipment to survive a MSLB inside containment will be performed after receipt of the applicant's submittal on this subject.
1
. NRC Response to Interrogatory 17.c.
Peter C. Hearn of the staff was involved in the evaluation of the Midland MSLB accident analysis. The staff's final report on this issue is contained in Section 6.2.1 of the Midland Safety Evaluation Report.
NRC Response to Interrogatory 17.d.
No.
NRC Response to Interrogatory 17.e.
Not applicable; see response to interrogatory 17.d.
Interrogatory 18 Contention 39 deals with the inadequacy of of Appendix J to set forth clearly the requirements for acceptable containment leak testing programs and for field inspectors to judge the acceptability of a licensees containment leak testing practices.
a.
What improvements have been made in Appendix J since 19787 provide documents that describe them.
b.
How does this problem specifically~ apply to the Midland nuclear plants?
c.
In their first sumary letter on Midland, the ACRS stated that l
B&W reactors have a higher leakage rate than other similar type reactors.
Provide documents on the extent of this higher leakage rate as compared I
to other reactors.
d.
What leak testing programs for the Midland nuclear plants has the staff found acceptable? Do the field inspectors agree that this is l
an acceptable leak testing program? Provide documents to demonstrate these facts.
e.
Has any staff member, field inspector, contractor or consultant disagreed as to the acceptability of the containment leak testing l
program? If so, provide documentatAon en the nature of their dissent.
l f.
If no improvements have been made, does Appendix J remain the regulatory requiremeat?
NRC Response to Interrogatory 18.a.
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The only improyemen',; to Appendix.J since 1978 was made in 1980 when 4
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i containment air locksimarcended. The amendment prescribes leak
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NPC ReLu)nse to Interrogatory 18.b.
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i ingperceived problem is that Appendix J d6es not clearly set forth
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i Appendix'J to mean 4. hat the Decay Heat Removal System and Service Water f
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It has been the Stoff's irterpretation that
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e these y'aMs must be corisidered for Type C testing;,bu't may be excluded i
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. not testing these valves. Consequently, the requirements of Appendix J were applied in a consistent manner, even though some discourse with the Applicant was needed to clarify the Staff's position.
NRC Response to Interrogatory 18.c.
NRC staff objects to this interrogatory. The subject of the contention is containment leak testing. The interrogatory addresses leakage rates from reactors, not containments.
NRC Response to Interrogatory 18.d.:
The leak testing program described in Section 6.2 of the Midland Final Safety Analysis Report and Section 6.2. of the SER is the program that the Staff has found acceptable.
The NRC Staff has not reviewed the Midland Integrated Leak Rate Test (ILRT) procedure; and, to our knowledge, the procedure is not ready for the staff's review.
NRC Response to Interrogatory 18.c.
No.
NRC Response to Interrogatory 18.f.
With the improvement involving the containment air lock testing, Appendix J continues as the regulatory requirement.
Interrogatory 19 Contention 40 deals with the lack of adequate qualification methods to satisfy the requirements for safety-related equipment established in IEEE standard 323-1974 for nuclear generating plants.
a.
Have adequate qualification methods been established to meet the IEEE Standard 323-1974 for safety-related equipment at the Midland 1 & 2?
b.
If the answer is yes, provide documents to substantiate your answer.
. c.
If the answer is no, what criteria for qualification and what standards are being used for safety-related electrical equipment?
Provide documents to substantiate your answer.
d.
Have any staff members or consultants disagreed on the adequacy of your methods or criteria for qualification of safety-related equipment? Provide documents that indicate the nature of their dissent.
NRC Response to Interrogatory 19.a.
Midland 1 & 2 are not required to meet IEEE Standard 323-1974.
NRC Response to Interrogatory 19.b.
Not applicable; see response to interrogatory 19.a.
NRC Response to Interrogatory 19.c.
The date of the Safety Evaluation Report for the Construction Pennit (CP-SER) determines which version of IEEE 323 is applicable to the electrical equipment in a particular plant. Since the Midland 1 & 2 CP-SER date is November 12, 1970, the 1971 version of IEEE 323 is applicable. The Category 11 positions of NUREG-0588 in conjunction with IEEE 323-1971 specifies the current requirements. A copy of NUREG-0588 is provided.
NRC Response to Interrogatory 19.d.
t No.
Interrogatory 20 Contention 41 deals with the present practices of permitting the connection of non-safety loads and required safety loads to Class IE power sources.
a.
Have any improvements been made in the manner in which non-safety loads and required safety loads and their connection to Class IE power sources since this contention was written in 19787 b.
If the answer is yes, provide documents to explain changes.
i
l
. c.
If the answer is no, explain the sequence of events that can happen with these types of connections that could lead to significant releases of radioactivity to the environment in the event of an accident, d.
What could be done, if anything, to intercept this course of events?
l e.
How does this problem specifically apply to Midland?
f.
List the specific measures that will be taken to resolve this issue.
NRC Response to Interrogatory 20.a.:
No.
NRC Response to Interrogatory 20.b.:
Not applicable.
NRC Response to Interrogatory 20.c.:
In the Midland design the non-safety loads are automatically disconnected from the Class IE power sources on detection of a safety injection signal. Since the non-safety loads would already be isolated from the Class 1E power sources under accident conditions, the staff is satisfied that these loads would not pose a threat to these sources.
Therefore, this design precludes a sequence of events resulting from non Class 1EloadsonClass1Ebusesthpcouldleadtosignificantreleases of radioactivity to the environment in the event of an accident.
NRC Response to Interrocjatory 20.d.:
As stated in response 20.c.,'the staff does not believe that the course of events will happen.
NRCResponsetoInterbogatory20.e.:
' :.- a
/
The Midland design includes the connection of both safety and selected non-safety loads to the Class 1E ESF buses of the ac onsite power systems. The connection of these non-safety loads to the Class 1E
. buses does not exceed the continuous rating of the emergency power sources.
In assuring that the interconnections between non-Class IE loads and Class 1E buses will not result in the degradation of the Class IE system, the isolation devices (circuit breakers) through which standby power is supplied to the non-Class 1E loads are tripped on detection of a safety injection signal. Therefore, the staff finds this design to be acceptable for the reasons stated in Response 20.c. above.
NRC Response to Interrogatory 20.f.:
The staff considers this issue resolved; therefore, no actions are necessa ry.
~
Interrogatory 21 Contention 42 deals with the fact that there is no assurance of adequate overpressure protection at Midland.
a.
Describe and document all the incidents of pressure transients in B&W reactors which have exceeded pressure temperature limits of the reactor vessels, b.
How were each of these incidents initiated? How were they resolved?
c.
How does this problem specifically apply to Midland?
d.
What specific measures are being taken to solve this problem at Midland?
NRC Response to Interrogatory 21.a Pressure temperature limits were exceeded at 1) Oconee Unit 2 on November 5,1973 during low power testing (The test procedures incorrectly instructed operating personnel to increase reactor system pressure in excess of the pressure - temperature limits.); 2) Rancho Seco on March 20, 1978 by oversupplying the steam generators with feedwater;
1 ~
- 3) TMI-2 on March 28, 1979 by refilling the Reactor System with cold ECCS water after the pressurizer relief valve stuck open and was subsequently isolated; and 4) Crystal River 3 on February 26, 1980 by continued inflow of cold ECCS water after the pressurizer relief valve control system failed causing the valve to open.
Documentation for these events is as follows:
1.
Oconee Unit 2
" Staff Discussion of 15 Technical Issues Listed Attachment to November 3,1976 Memo from Director, NRR, to NRR staff" NUREG-0138, November 1976.
2.
Rancho Seco - Letter from J. Mattimoe, SMUD, to R. Engelken, NRC, " Reportable Occurrence 78-1," March 31,1981.
3.
TMI-2
" Investigation into the March 28, 1979 Three Mile Island Accident by Office of Inspection and Enforcement,"
NUREG-0600, August 1979.
4.
Crystal River Letter from H. Denton, NRC to All Operating B&W Licensees, March 6, 1980.
These documents are provided.
NRC Response to Interrogatory 21.b.
The Oconee event was initiated intentionally by the operator in accordance with the existing test procedures, reactor system pressure was increased when the reactor system was at low temperature. The reactor system was returned to within the pressure-temperature limits following the test.
The Rancho Seco and Crystal River events were initiated by failure of the integrated control system and non-nuclear instrumentation. The Rancho Seco event was concluded when the operator stopped overfeeding the steam generators and depressurized the reactor system. The Crystal River event was concluded when the operator throttled ECCS flow and depressurized the reactor system. The TMI-2 event was initiated by a
l
. failure in the condensate system. Cleanup from the event is still in progress.
l NRC Response to Interrogatory 21.c.
Overpressurization at low temperature during testing does not directly apply to Midland. The Midland plants are protected from excessive reactor system pressure during both normal and low temperature operation, including testing, as discussed in Section 5.2 of the SER.
The events at Rancho Seco, Crystal River 3 and TMI-2 are not directly applicable to Midland. The Midland plants are protected from transients which might produce low reactor system temperatures coupled with ECCS actuation producing high reactor system pressure as discussed in Section 5.5 of NUREG-0793. The sections discuss protective features installed at Midland which were not available at Oconee, Rancho Seco, Crystal River and TMI-2.
NRC Response to Interrogatory 21.d.
As discussed in Section 5.2 of the SER, Midland Plants are protected against overpressurization of the reactor system while at low temperature byanalarmandpressurereliefcappility. The setpoints of this equipment are set below the maximum pressure limit for low temperature i
operation.
As discussed in Section 5.5.of the SER the Midland Plants are protected against steam generator overfill by safety grade main feedwater and auxiliary feedwat$r termination on high steam generator level and by
'g.
additional auxilicry feedwater flow control. The plants are protected r
against a stuck open pressurizer relief valve by a safety grade automatic isolation system.
. Interrogatory 22 Contention 43 deals with the vulnerability of the Midland reactors to industrial (or other) sabotage.
a.
FEMA has already indicated that the Midland area is a military target because of the Dow production and research operations here, some of which have military uses. Have precautions for security of the plant taken this fact into consideration?
b.
To what extent will the civil rights of people working in the nuclear plant, the Dow facilities and the community as a whole be violated as a means of security protection? This includes wire-tapping, surveillance, and other types of invasion of privacy.
c.
Has the public or the employees been advised as to how their civil rights will be affected in order to provide security for the Midland nuclear plant?
Response
Interrogatory 23 Contention 44 deals with the need to reexamine The Dow Chemical Co.
power systems as set forth in NUREG-0305 because of serious safety-related concerns.
a.
Has this reexamination of Dow Chemical power systems taken place for Midland?
b.
If the answer is no, how do you intend to compensate for this problem at Midland?
c.
List the specific measures that are being taken to solve this problem.
Response
Interrogatory 24 Contention 45 deals with the fact that the offsite power system for the Midland facility fails to meet the requirements of General Design Criterion 17.
. a.
Document the specific manner in which the offsite power source will interface with the onsite power systems at Midland, b.
Will any of this interaction depend on electrical equipment that j
has been stressed by the soil settlement proMem?
l l
c.
What are the special testing procedures that will be undertaken prior to operation to solve this problem at Midland given the unique soil settlement problems and their effect of unduly and unevenly stressing t
underground installations which includes electrical equipment?
NRC Response to Interrogatory 24.a.:
AsstatedinSection8.2oftheSER(NUREG-0793)thedesignofthe offsite power system for the Midland Plant meets the requirements of GDC 17. The offsite power system has two separate and independent circuits from the transmission network to the onsite distribution buses.
These circuits provide adequate capacity and capability to supply all station auxiliary loads as well as start and operate all safety-related equipment.
In addition, the offsite power system has sufficient electrical and physical independence such that no single event is likely to cause a simultaneous outage of both circuits to the onsite power distribution system. Each of these circuits would be available in I
sufficient time to prevent fuel damage and design conditions of the l
reactor coolant pressure boundary from being exceeded. This is I
consistent with the requirements of GDC 17.
The offsite power system interfaces with the Class IE onsite power systems through non-Class 1E 4.16 kV buses. Two separate and independent circuits from the transmission network connect the Midland Units 1 and 2 onsite distribution system to the station switchyard through startup transformers 1A and IB. Offsite power to the 4.16 kV redundant and independent Class 1E buses is provided through these two startup
. transformers, each having two secondary windings. One of the windings feeds a non-Class 1E 4.16 kV bus from which power is fed to its associated Class 1E 4.16 kV bus, and that other winding feeds a 6.9 kV non-Class 1E bus.
NRC Response to Interrogatory 24.b.
Settlement of the auxiliary and turbine buildings has been reviewed by the Staff and determined to be within the design allowables. Although it is still being evaluated by the applicant, the two buildings are believed to have maintained their structural integrity under soil settlement. Since the electrical equipment housed inside the buildings is much more flexible than the localized area of the structures to which it is supported or attached, small amounts of structural deformation will l
result in insignificant stress problem for the electrical equipment.
For the diesel general building, electrical equipment was not installed until the removal of the surcharge; therefore, no electrical equipment has been stressed by the soil settlement.
Finally, equipment seismic.and dynamic qualification is an on going program which is being reviewed by e Staff. Response spectra which are used for the qualification have been revised as a result of the soil settlement. The adequa6y of the applicant's equipment seismic and dynamic qualification program will be addressed in Section 3.10 of a supplement to the SER.
hRCResponsetoInterhogatory24.c.:
c.
The Staff has evalusted the adequacy of the Electrical Duct Banks following the differential settlement of the Diesel Generator Building.
The Electrical Duct Banks, which run under the Turbine Building from the l
. Diesel Generator Building to the Auxiliary Building, have been designed to assure that the cables within the.m remain functional despite the imposition of seismic and other loads. They are not, however, required to maintain a pressure boundary. Consumers Power Company (the Applicant) re-analyzed the duct banks for a limiting case and has repeated the analyses for three changes in the limiting case. No adverse effects have been reported.
The Applicant's analyses to evaluate the effects of soil / building differential movement indicated that the reinforcement did not reach the yield stress.
In addition, the Applicant has used a device (a " Rabbit") to check the availability of the individual ducts within i
each Electrical Penetration Duct Bank.
In regard to the Diesel Fuel Oil Tanks, the Applicant has analyzed and monitored them for effects caused by the supporting soil. The Applicant has not reported any problem areas from the analysis and monitoring program. Staff believes that the results of the analysis and monitoring program indicate that any structural concerns should be dismissed.
Interrogatory 25 i
Contention 46 deals with the absence of acceptable szandards and criteria governing the management of heavy loads near spent fuel.
l a.
How does this problem specifically apply to Midland?
1 b.
What specific measures have been taken to improve methods for handling this problem at Midland since it was identified in NUREG-0410?
c.
Describe and document incidents where this has been a problem at other operating reactors.
. NRC Response to Interrogatory 25.a.
This was an unresolved safety issue that has been resolved by NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants." At Midland specifically there was a possibility of a cask drop accident affecting either the spent fuel pool or safety related equipment below the spent fuel cask storage oit.
NRC Response to Interrogatory 25.b.
For the cask handling problem at Midland, Consumers Power Company has provided a single failure proof crane designed in accordance with the guidelines of NUREG-0554, " Single-Failure Proof Cranes for Nuclear Power Plants." Refer to Section 9.1.5 of our SER for our evaluation of this crane and the basis for our acceptance of the crane.
For the overall heavy-load-handling systems Consumers Power Company has responded to NUREG-0612 and has met the interim action items identified in NUREG-0612.
Refer to Section 9.1.5 of the Midland SER for further information regarding the overall heavy-load-handling systems. As identified in Appendix C of the Mir. land SER, NUREG-0612 resolves this safety issue.
NRC Response to Interrogatory 25.c.
There have been 34 Licensee Event Report (LERs) that report crane incidents at operating plants. These events involved a partial drop (15 inches) of a reactor vessel head without hitting any object, a 3 inch drop of the vessel head on the vessel flange, drop of a core barrel and internals (6 feet), damage to fuel during refueling, damage to nearby equipment by crane hook, dropping of polar crane hook, crane overload, damage to new fuel storage racks, and damage to a control room roof deck.
. None of the incidents resulted in a release of readioactivity.
For further information and evaluation refer to Section 4 of NUREG-0612.
Interrogatory 26 Contention 47 deals with the lack of a radionuclide/ sediment transport model which has been field verified, a.
How does this problem specifically apply to Midland?
b.
Document the specific measures that have been taken to solve this problem at Midland.
NRC Response to Interrogatories 26.a. and b.:
Contrary to what contention 47 states, there are field verified models available which can be used to calculate concentrations of 1
nuclides in surface and ground waters resulting from assumed accidental releases. However, for the Midland plant a transport model was not required because the staff used a highly conservative approach in evaluating radionuclide transport and assessing and bounding resulting potential doses as discussed in section 5.9.4.5 of the Midland Final Environmental Statement and section 2.4.7 of the Midland Safety Evaluation Report.
h Interrogatory 27 Contention 48 deals with the lack of an adequate analysis by the NRC staff to design basis floods.
a.
What improvements have been made by the Staff on design basis floods as it applies to Midland since it was identified as a problem by the ACRS and in both NUREG-0410 and the Black Fox testimony?
I Control o'f flooding at Midland depends on the integrity of a b.
series of dams on the Tittabawassee River system. Are there any plans for continued monitoring of these dams to be assured of their integrity?
c.
What will be the series of events that will take place at the Midland nuclear plant if flooding takes place?
. d.
In the event that all the dewatering systems break down, because of power failure during flood conditions, what will be the affect on the operation of the Midland nuclear plant?
NRC Response to Interrogatory 27.a.:
In the Black Fox Testimony cited in Contention 48, the staff concluded that its present deterministic approach to selecting the Design Basis Flood was the preferred method (as opposed to a probabilistic approach). Using this approach, the evaluation of the Black Fox Station led to the conclusion that the plant was adequately designed to accommodate the Design Basis Flood. Accordingly the conclusions in the Black Fox SER were unaffected by the generic issue of " Design Basis Floods and Probability." The design basis flood for Midland was also determined by using a deterministic approach.
NRC Response to Interrogatory 27.b.:
In determining the Design Basis Flood for Midland, it was assumed that all four dams on the Tittabawassee River would be overtopped and would fail during a Probable Maximum Flood. Consequently, the safe operation of the Midland Plant will not be affected should all four dams on the Tittabawassee River fail. The staff therefore has not pursued the question of monitoring of these dams and is thus unable to state whether or not they are or will be monitored.
NRC Response to Interrogatory 27.c.:
The Midland Plant grade is at an elevation of 634 ft. This is three feet higher than a maximum stillwater flood elevation of 630.4 ft.
(rounded off to 631 ft.) during the Design Basis Flood. Winds however, could cause waves to runup higher than plant grade. To prevent this runup from entering safety-related structures, watertight doors or
. watertight barriers will be installed on entrances located below the elevation of calculated maximum wave runup. Emergency procedures will describe the actions to be taken to ensure that watertight barriers and doors are installed where required in a timely manner.
NRC Response to Interrogatory 27.d.:
Should any disruption occur in the electrical power supply, a back-up diesel generator will be available to supply power to the interceptor wells until normal power can be restored. During a Probable Maximum Flood (PMF), the action of wind blowing across the water would create waves that could cause water to runup on the plant yard. Seepage into the plant fill would be minimal because there would be no ponding on site.
Instead, water would splash over and then recede until the next water wave approached the site. Any water that splashed over would be conveyed to the cooling pond or back to the river by both natural drainage and by a sloping on-site drainage system. Since there would only be minor seepage into the plant fill during a PMF, groundwater levels would not rise significantly beneath safety related structures even if the dewatering system failed.
Interrogatory 28 Contention 49 deals with the fact thsat there is no assurance that the design and operation of safety-related water supplies will insure adequate operation of the systems in the event of extreme cold whether and ice build-up, a.
How does this problem specifically apply to Midland?
b.
List the specific measures that are being taken to resolve this issue.
. NRC Response to Interrogatory 28.a.:
The only safety related water supplies that are located outdoors are the ultimate heat sink (UHS) and the borated water storage tank (BWST).
NRC Response to Interrogatory 28.b.
Protection against UHS freezing is evaluated in Section 2.4.5.2 of the Midland SER. During the winter months the VHS will be in use during normal and emergency operations and heated water will be discharged into the cooling pond so it is not expected that significant amounts of ice will accumulate on the water surface. Any surface ice formation would not affect UHS operation because the water entrance is about 34 feet below the normal water level of the cooling pond. With regards to ice loads, the cooling pond is designed to withstand the ice and static loads 5
induced by a 30 inch layer of ice in the event the plant were not in operation.
A separate tank heating system is provided for each BWST. The heaters are designed to maintain their structural integrity after a seismic event but are not required to maintain their heating function.
Heating of the BWST is not a safety elated function since the Technical Specifications require that the plant be shutdown following a BWST heater l
failure and there is sufficient time available to allow for safe plant l
shutdown following a heater failure. Also the chemical addition system which is located in a heated environment acts as a backup for plant shutdown.
,/
Interrogatory 29 Contention 50 deals with the fact that occupational radiation exposure to station and contractor personnel has been increasing, leading to hiring of transient workers which can increase the risk of operator
. error, sabotage, etc., as the Staff has recognized in NUREG-0410 and the Black Fox testimony.
a.
What methods have been taken to reduce occupational exposure at Midland since this probeim was identified in NUREG-0410 and the Black Fox testimony?
b.
Has any staff member or consultant disagreed with the adequacy of these measures?
c.
Describe and document the nature of these concerns.
d.
Are there plans to use transient workers at Midland at this time?
e.
If so, what kinds of criteria for qualifications of these workers will be used?
NRC Response to Interrogatory 29.a.:
With regard to implementation of occupational ALARA, the Midland management has provided a commitment to ensure that Midland will be operated in a manner consistent with the guidance of Regulatory Guides 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable", Revision 3, and 8.10, " Operating Philosophy For Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable."
The Midland plant was initially designed and reviewed using the ALARA policy, and the applicant and architect / engineer have continued to 1
l review, update, and modify the plant design during plant construction.
In order to reduce occupational radiation exposures to station and contractor personnel, the plant is designed to:
- 1) minimize the amount of personnel time spent in radiation areas; and 2) minimize radiation levels in routinely occupied plant areas and near plant equipment expected to require personnel attention.
In addition, operating and
. maintenance personnel at Midland will follow specific procedures to ensure that ALARA objectives are achieved in the operation of the plant.
Since the publication of NUREG-0410, the Staff has issued, for public comment, the draft version of NUREG-0761, " Radiation Protection Plans for Nuclear Power Reactor Licensees". This draft report provides guidance for NRC licensees and near-term licensees for the development and incorporation of a Radiation Protection Plan. The information contained in this report is the product of the NRC response to evaluations of the TMI accident, evaluation of industry-wide lessons learned, and significant findings derived from I&E's Health Physics Appraisals. This document will establish guidance and acceptance criteria for NRC Staff in determining the adequacy of power reactor radiation protection programs. One of the objectives of an effective radiation protection program is to keep radiation exposures to individuals with the limits of 10 C.F.R. part 20 and at levels which are as low as is reasonably achievable (ALARA). The establishment of effective radiation protection programs at nuclear power plants should result in an overall reduction in occupational exposures.
It is now the Staff intent to issue an amendment to NRC regulations to require that rodiation protection programs incorporate occupational ALARA rather than issue NUREG-0761 in its final form. When this regulation becomes effective, Midland will be required to implement program improvements, if any are necessary.
The applicant plans to minimize total occupational exposures at Midland by controlling quarterly doses to workers.
Pursuant to 10 C.F.R. 6 20.102, a licensee shall require any individual, prior to first entry
.._...m
. into a restricted area, under circumstances in which that individual could receive, in any period of one calendar quarter, an occupational dose in excess of 25 percent of the applicable standards specified in 10 C.F.R. 5 20.101(a) and 6 20.104, to disclose.in a written signed statement, either that the individual had no prior occupational dose during the current calendar quarter, or the nature and amount of such exposure. 10 C.F.R. 9 20.101 provides that, before pemitting any l
individual in a restricted area to receive a whole body occupational dose in excess of the standards specified in 10 C.F.R. 5 20.101(a), but within the limits of 10 C.F.R. % 20.101(b), each licensee shall obtain a certificate on form NRC-4 or signed statement from the individual containing all the information required in that form relative to such individual's accumulated occupational dose and other features outlined in the FSAR.
NRC Response to Interrogatory 29.b.
No.
NRC Response to Interrogatory 29.c.
Thisquestionisnotapplicablg.
NRC Response to Interrogatory 29.d.
i It is our understanding that Consumers Power Company will utilize contractors to provide skills and' services which are beyond the scope or capacity of Consumers Power Company.
NRC Response to Interrogatory 29.e.:
Contractorsuppliebworkerswillberequiredtohavequalifications to allow them to meet the qualification standards required for Consumers Power Company personnel.
. Interrogatory 30 Contention 51 deals with the fact that there is no assurance that existing geometry can adequately satisfy the functional design criteria for the behavior of fuel element assemblies during accident conditions, a.
How does this problem specifically apply to the Midland nuclear plant?
b.
List and document the specific measures that have been taken to resolve this issue.
NRC Response to Interrogatory 30.a.:
NRC Response to Interrogatory 30.b.:
Interrogatory 31 Contention 52 deals with the unreliable performance of diesel generators.
a.
Describe and document the incidents of failures in diesel generators at operating reactors.
b.
Provide documentation on the causes of these failures.
c.
How have these problems been resolved?
d.
Will the serious questions raised about the integrity of the diesel generator building itself further exacerbate the problems with the performance of the diesel generators? Provide documentation for the answer.
NRC Response to Interrogatory 31.a.
Incidents of diesel generator failure, as well as other types of incidents at nuclear power plants are documented in Licensee Event Reports (LERs). At the present time there is no published document which segregates LERs with respect to diesel generator fnilures only. LERs of all types (commencing Jan.-82) are listed in the monthly publication of NUREG/CR-2000,"LicenseeEventReport(LER) Compilation." This publication is the responsibility of the Office for Analysis and l
. Evaluation of Operational Data, U.S. Nuclear Regulatory Comission, Washington, D.C.
20555. A copy of NUREG-CR-2000 covering the May,1982 time frame is enclosed.
The following documents are known to present refined LER data on diesel generator perfonnance in varying formats:
" Data Sumaries of Licensee Event Reports of Diesel Generators at Nuclear Power Plants in USA" NUREG/CR-1362 03-80 "Comon Cause Fault Rates for Diesel Generators - Estimates Based on Licensee Event Reports at Nuclear Power Plants in USA" NUREG/CR-2099 Electric Power Research Institute (EPRI)
Publication NP2433 06-82 Copies of NUREG/CR-1362 and NUREG/CR-2099 are enclosed. For copies of EPRI NP 2433, contact the Electric Power Research Institute, Palo Alto, California.
NRC Response to Interrogatory 31.b.
Causes of failures of diesel generators, or other incidents at nuclear power plants, are documented in the Licensee Event Reports (LERs) submitted by the licensees. The availability of LER data is discussed in I
the NRC Response to Sinclair Interrogatory 31.a.
NRC Response to Interrogatory 31.c.
The University of Dayton Research Institute, under contract with the Nuclear Regulatory Comission, has conducted a study of diesel generator l
reliability at nuclear power plants. The results of this study are contained in a final report entitled " Enhancement of Onsite Emergency Diesel Generator Reliability," which was published as NUREG/CR-0660 in February, 1979. A copy of NUREG/CR-0660 in enclosed.
. In the report,10 areas are identified as significantly affecting diesel generator reliability. Specific reconinendations for corrective 4.ctions to improve performance in these areas and to improve overall diesel generator reliability are included in the report.
Following publication of NUREG/CR-0660, a letter was sent to all applicants for an operating license in which the applicants were asked to provide information on how the recommendations of NUREG/CR-0660 would be implemented. The Consumers Power Co. response to this issue is contained in Sections 9.5.4 through 9.5.8 of the Midland Units 1 and 2 Final Safety Analysis Report (FSAR). The Staff's findings on the acceptability of the information presented in the FSAR is documented in Sections 9.5.4 through
~
9.5.8 of the SER.
In summary, the Staff has concluded that the recommendations for improving diesel generator reliability, as stated in NUREG/CR-0660, have been or will be met or exceeded for Midland Plant, Units 1 and 2.
t I
NRC Response to Interrogatory 31.d.
Diesel generator performance, in general, is not affected by the structure in which it is located, egept for extremes such as total building failure, excessive differential movement between diesel generator and building foundations, or improper design of combustion air intake and exhaust systems.
For. Midland Plant, Units 1 and 2, there has been additional concern regarding diesel generator building settlement and the associated potential for damaging fuel oil and service water e-lines entering and exitirig the. building.
The Staff review of the applicant's remedial efforts on the seismic design of safety related structures for Midland Plant, Units 1 and 2, i
-B O
. including the diesel generator building, is not complete. However, the applicant's remedial efforts must result in a diesel generator building which conforms to NRC acceptance criteria and can withstand any design basis event without excessive differential movement between the foundations for the diesel generators and the diesel generator building.
When the Staff evaluation is completed, there should no longer be a question of building integrity affecting diesel generator perfonnance.
This subject is discussed in Sections 1.12, 2.5.4., and 3.8 of the Midland SER.
The Staff has concluded that diesel generator building settlement will not impair the structural integrity and functional capability of the underground diesel fuel oil and service water lines entering and exiting 3
the diesel generator building. The Staff has also concluded that the design of the combustion air intake and exhaust system is acceptable.
These items are discussed in Sections 3.9.3 and 9.5.8 of the SER, respectively.
Interrogatory 32 Contention 53 deals with the lack of adequate safety and environmental criteria for replacement of major pieces of equipment and of total decomissioning.
a.
What mode is now planned for decommissioning these plants?
l b.
If mothballing is the choice of decomission, who will pay for the guards, security, surveillance, monitoring and maintenance that the plants will require?
c.
If entombing in concrete is planned, have local and state officials been notified that because of the long half-life of nickel-59, which has a half-life of 80,000 years, means that the structures will l
remain there until they disintegrate and will have to be monitored permanently and with no tax base to pay for this?
. d.
How long can concrete structures already stressed by the soil settlement problems be expected to last?
e.
If dismantling will be done under water, where will the highly radioactive parts be stored?
f.
The costs of these options will vary greatly. How have they been considered in the cost benefit analysis? In what way? What are the guidelines for cost?
g.
What environmental and safety criteria have been established for the possible replacement of the steam generator or other major parts as has occurred at other nuclear plants?
NRC Response to Interrogatory 32.a.:
Consumers Power tentatively intends to decomission Midland, Units 1 and 2 through prompt removal / dismantling (Ref. 3).
In the NRC draft GEIS it was stated that this was the preferred method for decomissioning and when performed properly resulted in minimal impacts in terms of occupational dose and cost (Ref. 2). Other decomissioning alternatives were examined:
(1) safe storage followed by decontamination to unrestricted use and (2) entombment followed by surveillance and maintenance until radiation levels had decayed to unrestricted use levels.
It was concluded that in some situations the safe storage or entombment options were viable. However, in no situation was it considered acceptable to entomb a reactor with its internals, which contain long-lived radioactive activation products such as niobium-94 or nickel-59.
NRC Response to Interrogatory 32.b.:
Mothballing is not the initial choice for decomissioning Midland (Ref.3).
If it were, then provision for the cost of safe storage should be required as is recomended in the GEIS (Ref. 2).
. NRC Response to Interrogatory 32.c. and 32.d.:
Entombment is not the initial choice for decomissioning Midland (Ref.3).
If it were, then long-lieved activation products should require removal before such an alternative could be considered as is recommended in the Gels (Ref. 2). Moreover, the GEIS recomends that adequate funding should also be required.
In the GEIS it is indicated that if safe storage is preferred, then safe storage beyond 30 years from a health and safety viewpoint would be undesirable. Moreover, safe l
storage or entombment should be limited to no more than 100 years.
NRC Response to Interrogatory 32.e.:
The Battelle study addressed this problem (Ref. 1) along with all other problems required to safely decomission a PWR (using current technology) with a minimum of adverse impact. Prior to shipment offsite, the highly radioactive parts would be stored in the spent fuel pool.
Hnwaver, as was stated in the Midland response to this question, the dismantling plan should deal with storage and disposal of radioactive materials and take advantage of the then existing technology and disposal methods.
}
NRC Response to Interrogatory 32.f.:
The Consumers Power analysis did not specifically consider decommissioning cost-benefit. How'ever, decomissioning costs were estimated at $235 million for Midland, Units 1 and 2 for prompt removal /dismantlingib1984 dollars. As indicated in the Battelle J.-
e.
estimates (Ref. 2) decomissioning costs contain many site specific factors. However, for a rough comparison, the cost for decomissioning a gencric PWR was estimated. Based on the Battelle estimates, the Midland
o
. estimate should certainly cover the cost of decommissioning Midland, Units 1 and 2 using prompt removal / dismantlement and appears to be estimated on the high side.
In the GEIS (Ref. 2) it is recornended that proposed decommissioning rules require periodic updating of cost estimates to adequately account for various factors such as an expanding information base and improved technology, as well as fluctuations in general economic factors.
The GEIS included generic cost-benefit considerations and concluded that decommissioning considerations have a relatively small impact on overall cost-benefit considerations. Decommissioning costs were estimated as 5 to 10% of current dollar commissioning costs.
Public radiation dose from decommissioning activities were judged to be negligible and occupational dose was small in comparison to routine commercial occupational doses incurred during reactor operation.
For example, the prompt removal deconmissioning alternative resulted in an estimated annual dose of 300 person-rem / year over about four (4) years.
The average occupational dose for a pressurized water reactor, such as a Midland Unit, is about 500 person-rem / year over the estimated 30 years of reactor operating life. Of course, individual worker doses are always kept within limits of 10 CFR 20 which are limited to an annual dose of 5 person-rem / year.
NRC Response to Interrogatory 32.g.
The exposure criteria that the applicant will use for the possible future replacement of steam generators or other major components at the Midland plant are the same as used to maintain occupational exposures ALARA during normal plant operation. These criteria are based on the 1
guidance of 10 CFR 20 and the guidelines contained in Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As low As Is Reasonably Achievable", for maintaining occupation exposures ALARA.
References 1.
R. I. Smith, G. J. Konzek, and W. F. Kennedy, Jr., Technology, Safety and Costs for Decomissioning a Reference Pressurized Water Reactor Power Station, NUREG/CR-0130, prepared by Battelle Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Comission, June 1978. Addendum, published August 1979.
2.
Draft Generic Environmental Impact Statement on Decomissioning of Nuclear Facilities, NUREG-0586, U.S. Nuclear Regulatory Comission, January 1981.
3.
Affidavit of Louis S. Gibson, representing Consumers Power Company, th the U.S. Nuclear Regulatory Comission Atomic Safety and Licensing Board in answer to Mary P. Sinclair's Contention No. 53, July 1982.
l Interrogatory 33 Contention 54 deals with the possibility of damage to safety systems due to turbine missiles?
a.
The ACRS states that this plant is unusualy susceptible to the turbine missile problem. What additional safeguards will be provided to avoid this problem?
b.
Describe and document the incidents of turbine missile problems in operating reactors.
NRC Response to Interrogatory 33.a.:
The applicant has made an evaluation of the turbine missile risk for Midland Plant Unit Nos. I and 2.
Based on their analysis, which uses General Electric calculated probabilities for the generation of missiles l
from design and destructive overspeed failure of 8.7 x 10-9 per year and l
-9 5.0 x 10 per year, respectively, the probability of unacceptable damage l
l for Unit 1 is 1.4 x 10-9 per year and that for Unit 2 is 1.5.x 10-9 per year.
The applicant contends that their turbine wheel and overspeed protection system inspection and test programs are either ' explicitly or impIicitly incorporated in their evaluation and justify their use of the General Electric missile generation probabilities.
It is the staff's position that the relevant General Electric analyses be submitted io the staff for review and acceptance in order to verify the adequacy of the applicant's turbine inspection and test programs, and to demonstrate to our satisfaction a probability of unacceptable damage which is less than or equal to 10-7 per turbine year.
Ger.eral Electric has ne submitted reports to the NRC for' review which (a) incorporate stress oorrosion cracking as a turbine wheel failure mechanism in the analysis of missile generation probabilities, and (b) relate missile generation probabilities to turbine wheel and overspeed protection system inspection and test intervals. When their reports are completed and submitted to the NRC they will be reviewed and inspection and test intervals will established to meet NRC criteria for the probability of unacceptable damage due to turbine missiles.
NRC Response to Interrogatory 33.b.:
There have been three. incidents of turbine disk failures in operating nuclear reactors, only one of these resulted in missiles perforating the turbi,5e casing.
ThefollowingiIabrief_,descriptionofeach:
1.
Hinkley Point A (Ref. 1); Somerset, United Kingdom. On.
September 19, 1969 three disks on a low pressure. turbine failed during an 0
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}: _,
-)
~
(.-
~
- i
/
overspeedtr,iptestat3200 RPM (lGlIofratedspeed). Spontaneous w
[/f,
brittlefractureofIshrunk-onlowpressureturbinedisk,duetostress corrosion cracking in the crown of a keyway in the disk bore, initiated
'y the event. Three large segments of the fractured disks penetrated the
')
turbine c8 1ng. One of these missiles appeared to have passed through a joie in the casing'jnade by a prior missile penetration. The turbine was manufactured,Mbf the knglish Electric Company.
i 2.
Shippi,ngport(Ref.2);Shippingport, Pennsylvania. On
,g February 4,1974 the No.,2 low prqssure turbine disk burst during normal operation atil800 RPM. Mettalurchicalexsminationoftheburstdisk
-r fragments suggests that either stress corrosion or corrosion fatigue or both caus6d the crack growth that lead to brittie fracture. The
=
+
resulting fragments of tne fractured disk did not penetrate the turbine casing, consequently, there were no turbine missiles. The turbine was manufacturued by Westinghouse Electric Corporation.
1 Yankee Atomic Electric (Ref. 3); Rowe,' Massachusetts. On 3.
February 14, 1980 two of the disks in the low pressure turbine failed catastrophically during start up at 1800 RPM.
It is believed that one disk fractured into six sections, causing a series of events which resulted in the fracture of the other disk into two large sections and six small pieces. The cracks responsible for the initiating failure originated at the disk bore surface and are believed to have been due to stress corrosion. None of the broken pieces or disk sections penetrated the turbine casing, consequently there were no turbine missiles. The turbine was manufactured by Westinghouse Electric Corporation.
O s
References:
1.
D. Kalderon, et. al., Proc. Inst. Mech. Eng., Vol 186,31/72,1972, P. 341.
2.
Docket No. STN50-437-Reponse to Request for Information Concerning the Shippingport Disk Failure, letter P.B. Haga to K. Kniel, 9/23/76.
3.
A. Goldberg, and R.D. Streit, " Observations and Coments on the Turbine Failure at Yankee Atomic Electric Company, Rowe, Massachusetts," NUREG/CR-1884, UCID-18850, March 1981.
Interrogatory 34
~
Contention 55 deals with questions of adequacy of seismic design.
a.
Will the most recent seismic criteria be implemented at the Midland site for all the buildings when the plant begins operation?
b.
Which buildings will not be included?
NRC Response to Interrogatory 34.a.:
We assume that this interrogatory refers to the Site Specific Response Spectra when addressing "the most recent seismic criteria." The j
underpinning structures for the Auxiliary Building and the Service Water Pump Structure will be designed to the requirement of the Site Specific Response Spectra. The same applies for the foundation rings for the Borated Water Storage Tanks.
The existing structures will be evaluated for the Site Specific Response Spectra in a Seismic Safety Margin Evaluation.
{
I 1
(
O
, NRC Response to Interrogatory 34.b.:
All seismic Category I structures will be evaluated for the Site Specific Response Spectra. New structural components will be designed and built for these requirements while existing structures will be evaluated for Seismic Safety Margins.
Interrogatory 35 Contention 56 deals with the fact that Midland is not designed to accommodate a total loss of AC power.
a.
What back-up systems have been provided for loss of AC power?
Provide documentation.
b.
How will loss of AC power affect the operation of this plant?
c.
Describe and document incidents of the loss of AC power and their effects at other operating reactors.
NRC Response to Interrogatory 35.a.
The turbine driven auxiliary feedwater (AFW) pump has been designed to operate with a complete loss of AC power sources. Refer to Section 10.4.9 of the Midland SER for an evaluation of this capability.
NRC Response to Interrogatory 35.b.
J k
A loss of all AC power will result in reactor trip and initiation of the turbine driven AFW pump to supply water to both steam generators.
Decay heat will be removed by lifting of the steam generator safety valves and the adding of water from the condensate stroage tank via the turbine driven AFW pump. Thus, the plant will remain in hot standby.
.t NRC Response to Intbrrogatory 35.c.
As reported in the March 2, 1982 report from Oak Ridge National Laboratory there have been three documented' cases of complete loss of all AC power sources. They occurred at Haddam Neck (4-17-68), Lacrosse
. (9-17-74), and San Onofre (6-7-73). The Lacrosse and San Onofre events lasted 2 minutes and 1 minute, respectively, and both reactors were in a shutdown condition at the time. Therefore no adverse effects occurred.
TheHaddamNeckplant(ConnecticutYankee)wasat100%poweratthetime of occurrence. The offsite power was lost for 25 minutes. The event report is unclear as to how long both diesel generators were out of service (lessthan21 minutes). However, the Haddam Neck plant is equipped with two turbine driven pumps that require no AC power to deliver water to the steam generators for decay heat removal. No adverse effects were reported.
Interrogatory 36 Contention 57 deals with the fact that the electrical system will not function adequately under accident and/or fire conditions.
a.
What fire tests have been done since those made in September and October,19787 Document, b.
What improvements have been made in electrical wiring equipment since the September, October,1978 fire tests?
c.
How will the environmental qualifications be met for operating under accident conditions for the electrical equipment in the following critical safety systems: Containment spray, core flood, emergency core cooling, auxiliary feedwater, nuclear service water, containment isolation, decay heat removal and containment cooling.
NRC Response to Interrogatory 36.c.
IEEE 323-1971 and the Category II positions of NUREG-0588 set forth the criteria and requirements to which the equipment in the systems me
- <<, listed above must be qualified. Upon the receipt of Midland's final submittal on equipment qualification, the staff will review it against these criteria and requirements.
Respectfully submitted, William D. Paton Counsel for NRC Staff Dated at Bethesda, Maryland this 20th day of July 1982 s
i s
, - -, - -, w.
..o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of CONSUMERS POWER COMPANY Docket Nos. 50-329 'OL 50-330.0L-(MidlandPlant, Units 1and2)
CERTIFICATE OF SERVICE I hereby certify that copies of."NRC STAFF PARTIAL RESPONSES TO INTERROGATORIES SUBMITTED BY INTERVENOR SINCLAIR TO THE NRC STAFF ON JUNE 18, 1982" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or -as indi.cated by an asterisk through deposit in the
's Nuclear Regulatory Commission's internal mail system, this 28 day of July 1982:
- Charles Bechhoefer, Esq.
Frank J. Kelley Administrative Judge Attorney General of the State Atomic Safety and Licensing Board of Michigan U.S. Nuclear Regulatory Commission Steward H. Freeman Washington, D.C.
20555 Assistant Attorney General Environmental Protection Division Ralph S. Decker 525 W. Ottawa St., 720 Law Bldg.
Administrative Judge Lansing, Michigan 48913 Route #4, Box 190D Cambridge, Maryland 21613 h
Ms. Mary Sinclair 5711 Summerset Street Dr.' Frederick P. Cowan Midland, Michigan 4B640 Administrative Judge,
6152 N. Verde Trail Michael I. Miller, Esq.
Apt. B-125 Ronald G. Zamarin, Esq.
Alan S. Farnell, Esq.
Boca Raton, Florida 33433' Isham, Lincoln & Beale
- Dr. Jerry Harbour Three First National Plaza Administrative Judge 42nd Floor Atomic Safety and Li, censing Board Chicago, Illinois 60603 U.S. Nuclear Regulatory"Comission James E. Brunner, Esq.
Washington, D.C.
20555 Consumers Power Company
~
212 West Michigan Avenue Jackson, Michigan 49201
2-Ms. Barbara Stamiris
- Atomic Safety and Licensing Board 5795 N. River U.S. Nuclear Regulatory Commission Freeland, Michigan 48623 Washington, D.C.
20555 James R. Kates
- Ato ic S f ta e y and Licensing Appeal m
203 S. Washington Avenue Panel Saginaw, Michigan 48605 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Wendell H. Marshall, President Mapleton Intervenors
- Docketing and Service Section RFD 10 Office of the Secretary Midland, Michigan 48640 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Wayne Hearn Steve J. Gadler, P.E.
Bay City Tines 2120 Carter Avenue 311 Fifth Street St. Paul, MN 55108 Bay City, Michigan 48706 Frederick C. Williams Paul C. Rau Isham, Lincoln & Beale Midland Daily News 1120 Connecticut Avenue, NW 124 Mcdonald Street Washington, D.C.
20036 Midland, Michigan 48640 Myron M. Cherry, p.c.
Peter Flynn, p.c.
Cherry & Flynn Three First National Plaza Suite 3700 Chicago, IL 60602 T. J. Creswell Michigan Division Legal Department Dow Chemical Company Midland, Michigan 48640 0
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Willism D. Paton Counsel for NRC Staff m
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