ML20127M038
| ML20127M038 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/29/1992 |
| From: | Rogers K NRC COMMISSION (OCM) |
| To: | Steinbring G CITIZENS AWARENESS NETWORK |
| Shared Package | |
| ML20127C376 | List: |
| References | |
| NUDOCS 9301280088 | |
| Download: ML20127M038 (2) | |
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!Tx UNITED STATES g
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NUCLEAR REGULATORY COMMISSION IEs, v /,/
WASHINGTON. D.C. 20F46 December 29, 1992 CHAIRMAN Ms. Gail D. Steinbring President Citizens Awareness Network P.O. Box 83 Shelburne Falls, Massachusetts 01370
Dear Ms. Steinbring:
On behalf of the Commission, I am responding to your letter of November 2, 1992, in which you raised a number of concerns about the Yankee Nuclear Power Station and about the Nuclear Regulatory Commission's interactions with the nuclear industry.
The issues of whether Yankee Atomic personnel pressured the NRC staff to approve very relaxed " restart criteria" for the Yankee Rowe facility and relaxed embrittlement standards for the nuclear industry as a whole, and whether Yankee Atomic was allowed to edit Generic Letter (GL) 92-01 in such a way as to delete any reference to Yankee's attempts to minimize NRC scrutiny of reactor vessel embrittlement have already been addressed by the NRC staff in a letter dated June 22, 1992, to the Union of Concerned Scientists.
I am enclosing a copy of the letter for your information. The Commission is satisfied with the staff's handling of these matters.
With respect to the decision by the NRC staff to permit Yankee to terminate its off-site emergency plan, the staff's review of this action was very extensive and included five rounds of requests for additional information related to accident analyses. The staff not only reviewed credible accidents related to Yankee Rowe in its shutdown and defueled status but also examined the effects of releases beyond design basis events; for example, releases resulting from postuleted water loss from the spent fuel pool. During the Systematic Evaluation Program of older plants, completed for Yankee in 1987, the staff confirmed that the seismic analysis of the fuel pool was adequate.
In the most recent revieu, the staff found that because of the time elapsed since shut down and the low density of fuel storage, the off-site consequences of even beyond-design-basis events would result in off-site exposures of less than the-Environmental Protection Agency Protective Action Guidelines.
In M
addition, the staff concluded that the Defueled Emergency Plan (DEP) provides C
an acceptable emergency program for the Yankee Nuclear Power Station in its O
permanently shut down status and reasonable assurance that adequate protective H
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measures can and will be taken in the unlikely event of a radiological accident at the facility. The DEP requires the licensee to maintain a substantial emergency preparedness staff and agreements with off-site organizations such as medical facilities and police and fire departments until such time as the reactor fuel is removed from the spent fuel storage pool.
We have reviewed your comment "that the reactor is being allowed to carry on a wide range of decontamination techniques, many of which continue to release radionuclides into the Deerfield River."
The decontamination l
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, methods which have been used at Yankee have resulted either in solid waste products that cannot reach the river or in liquid wastes that are processed through the existing Radw=ste System in compliance with NRC requirements. The licensee maintained an eu.ellent record of compliance with NRC regulations regarding airborne and liquid releases to the environment while the plant was in operation. Now that the plant is shut down, all pre-decommissioning activities have been undertaken under the same radiological controls and gaseous and liquid effluent release criteria as those in effect when the plant was in operation.
The NRC recently reviewed the licensee's liquid effluent release data from July 1991 through November 1992, and concluded that liquid releases during this period were well below allowable limits. The NRC also reviewed the 1991 Radiation Environmental Program documentation and noted
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neither abnormal levels of radioactivity nor any adverse trends to the environment.
You also urged the NRC to completely address potential reactor vessel embrittlement problems at other operating nuclear plants in the United States.
We are already in the process of doing so. As a result of our evaluation of the Yankee Rowe vessel, the staff issued GL 92-01 to all reactor licensees.
This generic letter is only part of a larger ongoing program that will ensure that licensees are complying with existing regulations and other NRC guidance such as GL 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operation."
Regarding your request that we halt any further decontamination work at Yankee and hold full public hearings, you should be aware that in tha past four months, the NRC has issued six license amendments to Yonkee and has publicly noticed a seventh.
Each of these actions has provided the public an opportunity to intervene or comment. We have not received any comments or petitions to intervene. As required by our regulations, NRC also confers with the Commonwealth of Massachusetts State Liaison Official before the NRC staff issues such license amendmeni.s. We are considering a public meeting in the vicinity of the plant early in 1993 to provide information to the public on NRC's review of decommissioning in general and on expected site activities which will occur prior to the licensee's submittal of a decommissioning plan in late 1993.
1 hope these comments have helped to resolve your concerns.
If you have further questions, members of the NRC staff are available to discuss these issues in more detail.
As a starting point, you may want to contact Morton B. Fairtile, the NRC Project Manager for the Yankee Rowe facility, at (301) 504-1442.
Sincerely, M
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Kenneth C. Rogers Acting Chairman
Enclosure:
NRC letter of June 22, 1992
4 U CWSURE.
June 22, 1992 Ms. Diane Curran Harmon, Curran, Gallagher & Spielberg 2001 S Street, N.W.
Suite 430 a
Washington, DC 20009-1125
Dear Ms. Curran:
I am responding to your letters of April 30 and June 2,1992, in which you expressed concerns on behalf of the Union of Concerned Scientists, regarding the substantive contents of and the process for drafting Generic Letter (GL) 92-01, Revision 1, " Reactor Vessel Structural Integrity." The first concern expressed is that GL 92-01 appears to excuse widespread noncompliance. The second concern is that the U.S. Nuclear Regulatory Commission (NRC) apparently succumbed to inappropriate industry pressure in making revisions to the generic letter including providing a draft version of GL 92-01 to Yankee Atomic Electric Company (YAEC) without providing any equivalent opportunity for comment to the affected public.
In relation to the first concern, the NRC staff believes that licensees are in compliance with the regulations addressing reactor vessel structural integrity. The responses to GL 92-01 will provide information and data needed to confirm the NRC staff's understanding.
A detailed response to this concern is provided in Enclosure 1.
The second concern, that the NRC apparently succumbed to inappropriate industry pressure in making revisions to the generic letter and that a draft version of GL 92-01 was provided to YAEC, is incorrect.
The background portion of GL 92-01 was clarified and updated to better reflect YAEC extensive technical efforts regarding the fankee Nuclear Power Station reactor vessel integrity, thus assuring that the background information provided was fair and factual as detailed in Enclosure 2.
The enclosure also provides details of the process followed in making the changes and, as noted, the NRC staff has no evidence that YAEC was provided a draft of GL 92-01.
I trust that the information provided in the enclosures satisfactorily addresses each of your concerns and assists you in understanding the NRC staff's position on-the structural integrity of reactor vessels.
Sincerely, Gnf -
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. :M James M. Tayl'or :l
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/z. C Executive Director for Operations
Enclosures:
As stated cc w/ enclosures:
The Honorable Peter H. Kostmayer Response Cleared with Comissioners 6-22-92. !
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ENCLOSURE 1 RESPONSE TO THE CONCERN THAT GL 92-01 APPEARS TO EXCUSE WIDESPREAD NONCOMPLIANCE In your letter on behalf of the Union of Concerned Scientists of April 30, 1992, you requested that the NRC respond to the following questions:
1.
Which licensees are not in compliance with 10 CFR 50.00, 50.61, and Appendices G and H to 10 CFR Part 50?
The staff expects licensees to comply with these regulations. However, in view of certain concerns raised during the staff's review of reactor vessel integrity for the Yankee Nuclear Power Station (Yankee), the staff issued GL 92-01.
GL 92-01 is part of a program to evaluate reactor vessel integrity and take regulatory actions, if needed, to ensure that licensees and construction permit holders are complying with 10 CFR 50.60, 10 CFR 50.61, and are fulfilling commitments made in response to GL 88-11 "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and ii.s Impact on Plant Operation." Upon receiving the licensees' responses, we either will be able to confirm that all licensees are in compliance or will take appropriate action for any who are not.
In your letter you indicate that the NRC staff acknowledges in Enclosure 2 to GL 92-01 that only 16 of the country's 74 licensed pressurized water reactors have been found to comply with NRC's pressure vessel surveillance requirements in Appendix H to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50).
That is incorrect.
The portion of the letter you reference requested information from all addressees except those whose programs either meet American Society for Testing and Materials (ASTM) E 185-73, -79 or -82 or those who have an integrated surveillance program approved by the NRC, a group of 16 reactors identified in Enclosure 2 of GL 92-01. After issuing GL 92-01, the NRC staff found that the Crystal River Unit 3 Nuclear Generating Plant should also have been added to that list. The request for information from other addressees should not be used to infer that their surveillance programs are in noncompliance with NRC requirements.
During the licensing process, the NRC reviewed the programs of those facilities with invessel surveillance programs. However, many of the invessel surveillance programs were reviewed to earlier versions of Appendix H.
The NRC staff is concerned that the programs developed in accordance with ASTM E 185 editions earlier than -73 may not be effective in monitoring neutron embrittlement if the surveillance test samples do not contain all of the beltline materials or do not contain the limiting beltline materials. Therefore, in addition to responding to whether their surveillance programs are in compliance with Appendix H, licensees are being requested to explain how their surveillance programs will monitor neutron embrittlement if the programs do not contain the limiting base metal, weld metal, and heat affected zone material.
~ After issuing GL 92-01, the NRC staff learned that some licensees may have difficulty ensuring that their vessels comply with the Charpy upper shelf energy requirements of Appendix G to 10 CFR Part 50. The NRC staff will take appropriate action to ensure that licensees demonstrate that their reactor vessels will have margins of safety against fracture equivalent to those required by Appendix G of the ASME Code, or meet paragraph V.C. of l
Appendix G to 10 CFR Part 50.
I Appendix G to 10 CFR Part 50 requires reactor vessels to have Charpy upper shelf energies of 50 foot-pounds unless a lower value can be demonstrated j
to provide margins of safety against fracture equivalent to those required by Appendix G of the American Society of Mechanical Engineers Code. The Yankee licensee demonstrated that its reactor vessel could safely operate with a Charpy upper shelf energy of 35 foot-pounds.
Its analysis i
reaffirmed the conservative nature of the 50 foot-pounds requirement. As j
indicated, the NRC staff suspects that some licensees may have reactor vessels that are slightly below 50 foot-pounds.
However, the results of the Yankee analysis indicate that these licensees should also be able to f
demonstrate that their reactor vessels have the requisite margins of safety required by Appendix G to 10 CFR Part 50, 2.
Which licensees have failed to comply with the December 16, 1991, deadline for submitting safety analysis that justify exceeding the pressure vessel screening criteria?
On May 15, 1991, the NRC amended the Code of Federal Regulations at 10 CFR 50.61, " Fracture toughness requirements for protection against pressurized thermal shock events." The amendment changed the methods for calculating the pressurized thermal shock reference temperature (Rig) so that it is consistant with Regulatory' Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials." The amended rule states in part:
If the value of RT for any material in the beltline is projectedtoexcee5[,thePTSscreeningcriterionbeforethe expiration date of the operating license or the proposed expiration date if a change in the license has been requested, or the end of a renewal term if a request for license renewal has been submitted, this assessment must be submitted by December 16, 1991. Otherwise, this assessment must be submitted with the next update of the pressure-temperature limits, or the next reactor vessel material surveillance report, or 5 years from the effective date of this rule, whichever comes first.
The amended pressurized thermal shock (PTS) rule indicates that only those licensees who project that the RT value for their reactor vessels will exceedthePTSscreeningcriteriai.efore their licenses expire needed to submit assessments by December 16, 1991.
Each of the others is required to submit an assessment at a later date, as stated above.
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' The NRC staff has previously reviewed all reactor vessels to determine their compliance with the former version of 10 CFR 50.61. Based upon information supplied by licensees responding to the amended version of 10 CFR 50.61, the staff has determined that four plants could exceed the PTS screening criteria before the end of their current licenses:
Yankee Rowe, Palisades, Calvert Cliffs Unit 1, and Beaver Valley Unit 1.
All other pressurized water reactor vessels are predicted to be below the PTS screening criteria at the expiration of their licenses.
The vessel at Yankee Rowe is not of concern for PTS because it is no longer operating.
The licensees for Beaver Valley, Palisades, and Calvert Cliffs submitted assessments in accordance with 10 CFR 50.61. The NRC staff's interim safety evaluation for Palisades indicates that its reactor vessel will not reach the PTS screening criteria until well after 1995.
The licensee for Calvert Cliffs projected that the limiting weld in Unit I would exceed the PTS screening criteria in 2005 which is 9 years before its license expires. The licensee for Beaver Valley projected that one of its reactor vessel beltline plates would be 2 'F above the PTS screening criteria when its license expires in January of 2016.
The NRC staff is reviewing these submittals to determine the adequacy of the flux reduction measures and the other actions being taken by these licensees to address PTS effects on their reactor vessels.
Using previously-reported chemical composition of the beltline materials and projected neutron fluence at the end of current licenses, the staff performed an independent evaluation of licensees' reactor vessels in accordance with the methodology in the amended 10 CFR 50.61 to determine which vessels would be above the screening criteria at the end of their current license.
Based upon this evaluation, the staff concluded that all F
other reactor vessels are projected to be below the PTS screening criteria at the end of their current licenses. Therefore, the staff believes that all licensees who were required to comply with the December 16, 1991, deadline have done so.
3.
Why has the NRC staff excused compliance with this deadline?
The NRC staff has not excused compliance with the December 16, 1991, deadline.
4.
Has the staff granted a blanket extension of up to nine years for '
licensees who have failed to comply with Appendix H to 10 CFR Part 50?
If so, what is the justification for the extension?
The NRC staff has not granted a blanket extension for noncompliance with Appendix H.
The NRC reviewed many of the invessel surveillance programs to earlier versions of Appendix H and now considers it prudent to verify that these surveillance programs comply with the applicable version of ASTM E 185, in accordance with the 1983 version of Appendix H.
Thus, the NRC staff intends to assure that licensees responding to GL 92-01 are in compliance with Section ll.B of Apcendix H.
t
. Your letter also referenced the July 11, 1991, Commission briefing on Yankee, indicating that ".
licensees comply with Appendix H.". the NRC Staff expressed uncertainty as to whethe to whether tar licensees were in compliance with Appendix H.The uncertainty Rather the answer expressed uncertainty as to whether there were any other licensees who needed and had not requested exemptions from Appendix H following its revision in 1983.
The staff expects all licensees to be in compliance. Hosever, based upon the Yankee review, the staff considers it prudent to go back and verify that all licensees are in compliance.
In summary, the NRC staff believes that licensees are in compliance with the regulations addressing reactor vessel structural integrity as noted in the above responses.
The responses to GL 92-01 will provide information needed to confirm the NRC staff's understanding.
The NRC staff also believes that the regulations on the structural integrity of reactor vessels are sufficiently conservative so that reactor vessels approaching the criteria for ensuring compliance with the regulations retain an adequate margin of safety from potential reactor vessel failure.
ENCLOSURE 2 3
RESPONSE TO THE CONCERN THAT THE NRC APPARENTLY SUCCUMBED TO INAPPROPRIATE INDUSTRY PRESSURE IN REVISING GL 92-01 In your April 30, 1992, letter, you requested that the Commission give an
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accounting as to how an; why the revisions were made to GL 92-01.
The NRC staff revised the background information portion of GL 92-01 precisely for the reasons stated in the opening paragraph of GL 92-01, Revision 1, which states:
This letter replaces Generic letter 92-01 dated February 28, 1992.
g The background information concerning NRC's assessment of embrittlement in the Yankee Nuclear Power Station [ Yankee] reactor vessel was drafted by staff some months ago and has now been clarified and updated to better reflect the licensee's extensive technical efforts regarding reactor vessel integrity.
The section pertaining to required information has not changed.
As stated above and noted in your letter, the NRC staff took several months to prepare GL 92-01. During the final review of GL 92-01, some members of the NRC staff felt that the background portion should be revised as demonstrated by an internal memorandum of January 30, 1992, (Attachment 1) in which they proposed changes to the background portion.
The proposed changes were consideredandadecisionwasmadetoissueGL92-01withoutincorporatingth changes in order to issue the request for information to licensees as promptly as possible.
Yankee Atomic Electric Company (YAEC) obtained a copy of GL 92-01 which was issued on February 28, 1992.
Following its review of the letter, Mr. J. D. Haseltine, YAEC Project Director, called Mr. James Partlow of the NRC staff on March 2, 1992. Mr. Haseltine expressed concern about the completeness and negative characterization of the Yankee vessel evaluation in the generic letter. Mr. Partlow stated that if YAEC did not believe the information was fair and accurate, he should provide his comments to the staff and they would be evaluated. Mr. Haseltine telefaxed his written comments in a draf t letter to the NRC dated March 3,1992, (Attachment 2).
Since the letter was dated but not signed, the NRC staff believed the draft letter was an advance copy of a letter that would be submitted formally to the NRC.
On March 4, 1992, Dr. Andrew Kadak, President of YAEC, met separately with the Chairman; Commissioners Rogers, Curtiss, and Remick; and Dr. Thomas E. Murley to discuss various issues. During Dr. Kadak's discussions with Chairman Selin and with Dr. Murley, he raised concerns similar to those expressed in the March 3,1992, draft letter regarding the background portion of GL 92-01.
(GL 92-01 was not discussed with the other Commissioners.) As documented in a memorandum of April 14, 1992, (Attachment 3) summarizing Dr. Kadak's visit, the Chairman asked the NRC's Executive Director for Operations to ensure that the background information pertaining to Yankee Rowe was fair and factual.
In addition, following Dr. Murley's meeting with Dr. Kadak, Dr. Murley asked Mr. Partlow to look into the mattcp and ensure that the background information was fair and factual.
4 )
Following receipt of the March 3,1992, draf t letter from YAEC, senior members of the NRC staff evaluated the proposed changes to the background portion of GL 92-01. Those changes that the staff considered appropriate to ensure the accuracy and fairness of the background portion were included in Revision 1 to GL 92-01.
Although the staff evaluated and incorporated some of the changes proposed by YAEC, these changes were not made as a result of pressure from the licensee.
Rather, these changes were made in response to legitimate concerns raised by the licensee in order to ensure that the background section of GL 92-01 was fair and factual and reflected the licensee's extensive efforts to resolve issues regarding the Yankee reactor vessel.
You also requested the following information:
1.
Any written documentation of YAEC's request for changes to GL 92-01.
Attachments 2 and 3 contain this documentation.
2.
Identify all NRC personnel contacted by YAEC for the purpose of requesting changes to GL 92-01.
YAEC contacted Chairman Ivan Selin, Dr. Thomas E. Murley, and Mr. James G. Partlow.
3.
Supply any memoranda, notes, or telephone records of any contacts between the NRC staff and/or Commissioners and YAEC regarding changes to GL 92-01 that were requested by YAEC.
Attachments 2 and 3 provide the it. formation.
4.
Documents that explain the basis for the changes made by the NRC to GL 92-01.
The basis for the changes to GL 92-01 is stated in the opening paragraph 4
of Revision 1 to the generic letter and the process for making the changes are described in this enclosure.
In your June 2,1992, letter, you state that on May 13, 1992, the NRC, without any notice to UCS, placed in its Public Document Room (POR) two documents relevant to the process by which GL 92-01 was revised.
These documents were a March 3, 1992, draft letter to NRC from J. D. Haseltine, YAEC's Project Director, enclosing a marked-up draf t of GL 92-01.
From your review of the documents, you conclude that YAEC had received and edited a draft copy of GL 92-01.
1 e With respect to placement of Mr. Haseltine's March 3,1992, draf t letter in the POR, upon reviewing your April 30, 1992 letter, the staff realized that it had not placed Mr. Haseltine's letter in the POR.
Since this document aid form part of the basis for the changes to GL 92-01, it should have been placed in the POR. To correct this situation, the staff placed the document in the POR.
With respect to your concern regarding YAEC being provided a draft of GL 92-01, that conclusion is not correct. Although, Mr. Haseltine's March 3, 1992, letter refers to "the staff's draft Generic Letter," that statement is not correct. We have no evidence that YAEC was provided a draft of GL 92-01.
A copy was available to YAEC through the NRC's Generic Communications Electronic Distribution System (GCE05) on February 28, 1992.
GCEDS is an electronic mail system through which the text of NRC generic communications may be accessed upon their issuance.
This system is available at no cost to interested utilities, vendors, and members of the public.
Access to this system may be obtained through the NRC's Generic Communications Branch. Thus, the " annotated copy of the Generic letter" enclosed in Mr. Haseltine's March 3, 1992, draft letter is YAEC's retyped proposed revisions (sections added and deleted) to the background section of GL 92-01 and was based upon the as-issued generic letter and not a draft of GL 92-01.
Your June 2,1992, letter also stated your understanding that the Commission recently proposed to require the publication of draft generic letters in the Federal Register.
It asked if that understanding is correct and, if so, when the Commission intends to finalize this requirement.
On December 20, 1991, following its review of SECY-91-172, " Regulatory Impact Survey Report -
Final," the Commission directed the staff to propose a methodology to solicit the views of interested groups on proposed generic communications in which a new staff position is articulated or through which the staff seeks additional licensee commitments.
The staff has this methodology under development and is scheduled to submit its proposal to the Commission in early July 1992.
Both SECY-91-172 and the December 20, 1991, Staff Requirements Memorandum are available in the NRC's Public Document Room.
l' Attachnent 1 January 30,1992 MEMORANDUM FOR:
B.D.
LIAW, DEPUTY DIRECTOR,DET FROM:
J.T.
WIGGINS, ACTING CHIEF, EMCB
SUBJECT:
CONCERNS WITH PROPOSED GENERIC LETTER ON RPV STRUCTURAL INTEGRITY As we discussed on 1/29/92, I am concerned that the staff has been too categorical in its discussion in the proposed GL regarding YAEC's compliance to 10CFR 50.60, 50.61, Appendices G and H and GL 88-11. Although I do not question the adequacy of our staff's reviews or the technical conclusions reached, I believe that the language in the GL prematurely finds the licensee in violation of those requirements.
When a licensee is found to be in violation of our requirements, we typically address this matter formally through the Enforcement Policy. By using the Enforcement Policy, the licensee is afforded an opportunity to contest the proposed violation (s) and to comment on the safety significance of the issue at hand. Further, the licensee can address proposed corrective actions. This practice provides the licensee a form of due process.
By not using our normal methods to pursue potential violations, we leave ourselves open to a claim from the licenses of unfairness.
The disposition of that claim could serve to fog the central technical safety issue that the GL intends to address.
I've marked up the proposed GL to soften its statements regarding compliance to the regulations by stating that the staff has raised questions as opposed to concerns or findings. I'd ask that my mark-up be reviawed.
As a side issue, the question of enforcement will still need to be addressed; however, this is not the appropriate time. The staff and the Yankee licensee need to continue to focus their energies toward.
resolution of the RPV technical safety issues without being distracted by the legal ramifications of those issues.
TW Ja s T. Wiggins T
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A UNITED STATES j
>j NUCLEAR REGULATORY-COMMISSION j
g WASHINGTON. D.c. MM -
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j TO:
ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS POWER PLANTS (EXCEPT YANKEE ATOMIC ELECTRIC COMPANY, LICENSEE FOR THE YANKEE NUCLE *R POWER STATION).
SUBJECT:
REACTOR VESSEL STRUCTURAL INTEGRITY, 10 CFR 50.54(f)
(GENERIC LETTER 92-
)
The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to.
obtain information-needed to assess compliance with requirements and cosmitments-regarding reactor vessel integrity-in view of certain concerns raised in-the i
staff's review of-reactor vessel integrity for the Yankee ituelear Power Station.
In Section 50.60(a) of Title 10 of the Code of Federal Regulations'(10 CFR:
50.60(a)), the NRC requires that licensees for all light water nuclear power reactors meet. fracture toughness requirements and have-a material: surveillance; program for the reactor coolant pressure boundary.- These requirements are set-forth in Appendices G and'H to 10 CFR Part 50.
In 10 CFR 50.60(b),'where the-requirements-of Appendices G and-H to.10:CFR Part 50 cannot be met. an exemption -1 is necessary pursuant to 10 CFR 50.12.
In.10 CFR 50.61_the NRC also provided fracture toughness requirements for protecting pressurfred water reactors against pressurized thermal shock events.
Licensees and permit holders have -
also made commitments in' response to Generic. Letter (GL) 88 11, "NRC Position -
on Radiation Embrittlement of Reactor. Vessel Materials;and its Impact on Plant Operations," to use the methodology in Regulatery Guide 1.99, Revision 2,-
" Radiation Embrittlement of Reactor Vessel Materials," to predict the: effects of neutron irradiation as required by' Paragraph V.A of 10 CFR ' art 50, Appendix G.
The 10 CFR 50.60 and 10 CFR-50.61 requirements and-GL 881& are in the-
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overall regulatory program to maintain-the structural. integrity of the reactor-vessel. While reviewing the integrity of the reactor vessel at the Yankee-Nuclear Power Station, the NRC _ staff raised ramma=megegarding the licensee's compliance with certain requirements and commitments.L This generic-letter is part of a; program to evaluate reactor vessel integrity and take regulatory actions, if needed, to ensure that licensees and permit-holders are complying with 10 CFR 50.60 and 10 CFR 50.61, and are fulfilling-
.cosmitments made-in response to GL 88-11. is a discussion of'the applicable -regulatory requirements.
The NRC is re compliance under the provisions of 10 CFR 50,54(f) quiring information on Assessment of Embrittlement for the Yankee Nucleadower_ Station Reactor-vessel In an effort to resolve concerns regarding the neutron embrittlement of the Yankee reactor vessel, the staff performed a safety assessment of the Yankee 3
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reactor vessel. [The staff,-h iht the lice'nsee f or the Yankte Nuclear Power Station =fght not be-in compliance with 10 CFR 50.60 and/had se6 properly completed the assessment recuired in 10 CFR 50.61.
- Further, the Yankee Nuclear Power Station had incorrectly applied the,the licensee for thodology in Regulatory Guide 1.99, Revision 2.
g mgj The staff t the Charpy upper shelf energy of the Yankee reactor vessel material could be as low as 35.5 foot-pounds which is less than the 50 foot-pound value required in Appendix G to 10 CFR Part 50)the actic s m&:'
5.;rcu -the licensee for the Yankee Nuclear Power Station had not performed Q ragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
Since t en, the licensee nas 4 performed an analysis in accordance with Paragraph IV.
1 of Appendix G to 10 CFR Part 50 using criteria being developed by the Ame ican Society of Mechanical Engineers (ASME) to demonstrate margins of safety eq valent to those in the ASME Code.
wgg The NRC m ~u Lat y A M al
!!L _ m - regarding compliancc with the requirements of Appendix H to 10 CFR Part 50. Section E 185 of the American Society for Testing and Materials (ASTM) Code requires that the licensee take sample specimens from actual material used in fabricating the beltline of the reactor vessel. These surveillance materials shall include one heat of base metal, one butt weld, and one weld " heat affected zone."
The licensee for the Yankee Nuclear Power Station terminated the material surveill),nce gegram in 1965.4 3
Therefore,gthe Yankee Nuclear Power Station had MTruitemal ~ surveillance e proaram an July 26.1983. shen Annandh H rn in!rro part 50 became effective. _ j
- Further, ggg af base w)the samples irrayiated at Yankee Rowe;before 1965 were compfsed only '
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-g e licen r the Yankee Nuclear Power Station had used the methodology in y
Degulatory Guide 1.99, Revision 2 to predict the effects of neutron ertritt leme nt.
However, the staff found that the methodology in Regulatory Guide 1.99, Revision 2, = ! 4 correctly applied b The specific issues were (1) the irradia ion temperature, (2) y the licensee.
the chemistry composition of reactor vessel material, a (3) the results of the material surveillance
- program, g
The irradiation temperature at t e Yankee Nuclear Power Station is between 454 *F and 520 'F, which is below the nominal irradiation temperature of 550 'F used in developing Regulatory Guide 1.99, Revision 2.
A lower irradiation temocrature increases the effect of neutron eetrittlement. The regulatory guide indicates that for irradiation temperatures less than 525 'F, es6rittlement effects should be considered to be greater than predicted by theM **
methods of the guide.
Adjustments that were made by the licensee were g
,L h )dsufficient to account for this effect.
The limited results of the surveillance program from the Yankee Nuclear Power Station indicated that the increase in the reference temperature exceeds the mean-plus-two standard deviations as predicted by the procedures in Regulatory Guide 1.99 Revision 2.
The regulatory guide states that the licensee should usecrediblesurveillancedatatopredicttheincreaseinreferencetemperature resulting from neutron irradiation.
. The staff implemented RG 1.99, Revision 2, by issuing GL 88-11.
In committing to GL 88-11, licensees have comitted to calculate radiation embrittlement in accordance with the procedures documented in RG 1.99, Revision 2.
To meet the limitations in Section 1.3 of the regulatory guide, the licensee should consider the effects on irradiation embrittlement during core critical operation with irradiation temperatures less than 525 *F.
Section 2 of the regulatory guide states that the licensees should consider the effects of the results from its surveillance capsules.
The Suswer 1972 Addenda of the 1971 Edition of Section III of the ASME Boiler and Pressure Yessel Code are the earliest code requiremer.ts for testing materials to determine their unirradiated reference temperature.
Since the Yankee reactor vessel was constructed to an ASME Code earlier than the Susiner 1972, it had not been sufficiently tested to determine its unirradiated reference temperature.
The licensee for the Yankee Nuclear Power Station extrapolated the available test results to determine an unirradiated reference temperature.
The staff pt g :
ther licensee's extrapolation was gconsgative.
The chemical composition of the Yankee reactor vessel welds 1 N ": 3 The
+#
material's sensitivity to neutron embrittlement depends on its chemical content.p'L The licensee assuned that the chemistry of its welds was equivalent to that of g the BR-3 reactor vessel in Moi, Belgium.
However, the licensee could not identify the heat number of the wire used to fabricate the Yankee welds. The licensee was assuming a chemical composition that was not based on its plant-specific information, since the chemical composition, in particular, the amount of copper, depends upon the heat number of the weld wire.
to b thet-see for the Yankee These factors prompted th s Nuclear Power Station had considered plant-specific information in assessing comoliance with 10 CFR 50.61.
When4 plant-specific information,ii m i a n, J u l thf the Yankee reactor vessel may A4*e exceeded the, screening criteria in 10 CFR 50.61. Since then, the licensee has performed a probabilistic fracture mechanics analysis in accordance with 10 CFR 50.61(b)(4) and the staff is continuing its review.
Q M
Upon conducting the Yankee Nuclear Power Station review, the staff became concerned that this may not be an isolated case regarding compliance with 10 CFR 50.60 and 10 CFR 50.61 and fulfillment of commitments made in response to Gl. 88-11.
Thus, the staff is issuing this generic letter to obtain information to assess compliance with these regulations and fulfillment of commitments.
m with the Yankee Atomic Electric The staff is continuing to pursue this = : kie Company need not respond to Company. Therefore, the Yankee Atomic Elec this generic letter.
(
Reouired Information Portions of the following information requested is not applicable to all addressees. The responses provided should, in these cases, indicate that the requested information is not applicable and why it is not applicable.
I 1
mR.03 '9214:14 YRtE Riort1C DICTRIC CD/DOLTON P.BZ
\\
l YANKEE ATOMICELECTRIC COMPANY
'*'*y*l ATTACHMENT 2
%'7"l"
\\
3 580 Main Street. Botton, Massachusetts 017401398 i
i March 3, 1992 BYR 92-027 YPV 120/92 U. s. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attention:
Mr. 7ames G. Partlow j
Associate Director for Projects
Reference:
(a)
License No. DPR-3 (Docket No. 50-29)
Subject:
Conments on Generic Letter 92-01, Reactor Vessel structural Integrity
Dear Mr. Partlow:
We have carefully reviewed the staff's draft Generic It'ter 92-01 (Reactor Vessel Structural Integrity) dated February 28, 1992.
As I discussed with you in our March 2 telephone conversation, Yankee is very concerned about the incomplete and negative characterization of the Yankee vessel evaluation.
Enclosed is an annotated copy of the Generic Letter which we believe nere correctly portrays the history as well as the recent-interaction by the Staff and Yankee regarding the Yankee vessel.
Most of the annotations add additional information to reflect the knowledge gained over a period of time by both the Staff and Yankee.
It is simply not sufficient to cite that a methodology was incorrectly applied without an accompanying explanation as to how the analysis and decision making evolved as more and more knowledge was gained.
In some instances guidance evolved as analyses.on.the -
Yankee vessel were performed.
The correction for. temperature offoct is a good example.
The Generic Letter needs to make clear that the Yankee evaluations were in many ways a first-of-a-kind application and interpretation of certain methodologies and Regulatory Guides.
Our annotations also seek to establixn the fact that Yankee has performed high quality and professional work. As the letter is now written, there are several references to nethodologies being
... incorrectly applied by the license."
While there were reasonable disagreements between the Staff and Yankee engineers, phrases like the one cited above harus Yankee's reputatien and inappropriately ignores the learning curve that everyone was on.
I 1
'9205130170 920303
____[/d qDR ADOCK0500gj9
MAR.03.'92 14: 16 YFt+IE ATOMIC ELECTRIC CD/BOLTON P 03
(
- 0. S. Nuclear Regulatory Co=, mission March 3, 1992 i.
Page Two y,y i
I j
j t
i We believe. Dr.
Murley's comments on February 7th correctly l
characterized the intaraction.
I "Is integrity] resolved in a high class technical way."m glad be foraally incorporated by the staff into a final Gerieric Please do not hesitate to call if you have questions or require further explanation of our annotations.
Sincerely, J. D. Haseltine JDH/kg Project Director i
i i
rFIR.03 "3214: 17 YftCTE ATOr18C CLECTRICJ;0/DOLycN P.04 i
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TO:
ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FOR NUCLEAR POVER PLANTS (EXCEPT YANKEE ATOMIC ELECTRIC CCHPANY, LICENSEE FOR THE YANKEE NUCLEAR POWER STATION)
SUBJECT:
REACTOR YESSEL STRUCTURAL INTEGR!TY,10CFR50.54(f) (GENERIC LETTER 92 01)
The U.S. Nuclear Regulatory Corm:ission (NRC) is issuing this generic letter to obtain information needed to assess compliance with requirements and coccitreats regarding reactor vessel integrity in view of certain concerns raised in the staff's review of Reactor vessel integrity for the Yankee Nuclear power Station. In Section 50.60(a)5 of Title 10 of the Code of Federal Regulations (10CFR50.60(a)), the NRC requires that licensees for all light water nuclear power reactors meet fracture toughness requirements and have a material surveillance program for the reactor coolant pressure coundary. These requirements are set forth in Appendices G and H to 10CFR, Part 50.
In 10CFR50.60(b), where the requirements of Appendices G and H to 10CFR Part 50 cannot be met, an exemption is necessary pursuant to 10CFR50.12.
In 10CFR50,61 the NRC also provided fracture toughness requirements for protecting pressurir.ed water reactors against pressurized thermal shock events. Licensees and permit holders have also made cocrnitments in response to Generic Letter (GL) 8811, 'NRC Position on radiation Emorittlement of Reactor Vessel Hsterials and its impact on Plant Operations,"
to use the methodology in Regulatory Guide 1.99, Revision 2,
- Radiation Embrittlement of Reactor Vessel Materials," to predict the effects of neutron irradiation as required by Paragraph V.A of 10CFR, Part 50. Appendix G. The 10CFR50.60 and 10CFR$0.61 iequirer:ents and GL 8811 are in the overall regulatory program to maintain the structural integrity of the reactor vessel.
Wh'ic rivicwing the 'nicgrit3 ;; i,h; rcectcr e;;;ci et the YcoLcc4vekee Ice;i St;tica, thc CC ateff.,5sd L56 %..5 sciiiGir.s the licinacc'3 ce@encc.;ith ccriein icMic;;at; end cdt; cats.
This generic letter is part of a program to evaluate reactor vessel integrity and take regulatory actions, if needed, to ensure that licensees and permit holders are complying with 10CFR50.60 and 10CFR30.61, and are fulfilling commitments made in response to GL 8811. Enclosure 1 is a discussion of the applicable regulatory requirements. The NRC is requiring information on compilance under the provisions of 10CFR50.54(f).
R72\\49 1
I i
l.
rm. 03 ' 92 14:19 WetTE NTOr11C El.ECTRIC CD/DOLTm P.25 Assessment of Embrittlement for the Yankee Nuclear Power Station Reactor Yessel 5
in an effort to resolve concerns regarding the neutron embrittlement of the Yankee reactor vessel, the staff performed a safety assessment of the Yankee j reactor vessel for Ebc steff revic.cd fe nd thet the licanacc fer ;r; 7enkcc Mecicer re.cr Stetien sight ne; tc in compliance with 10CFR50.60 and hed nec rieptrly ce;pictcd th; ;eets;; cat rcq.ircd in-10CFR50.61.
Terttcr. the licenacc for thc 7enkti N cicer re.cr Statica hed inc;rrectli eppllcd the I
=cthedeiepy in ac;.1;ter, Oeidc 1.0^. Rc.iaien O.
't l Thc ;;;ff found that Yankee estimated that the Charpy upper shelf energy of l the Yankee reactor vessel material could be as low as 40 SSTS foot pounds for
] weld and 35 foot pounds for plate which is less tian the 50 foot-pound value l required in Appendix G to 10CFR, Part 50. In N6 REG 0569, the NRC staff l evaluated the low upper shelf conditions of the Yankee reactor vessel. It was l concluded at this time that Yankee plate and weld materials met the l requirements of Appendix G through the end of license. However, thi licensce fcr th; 7;nkc; "scicer I;-;r Stet 4cn had nei sciTeracd th; action; ic; ired in Per esreshi F.'. A. : er !.0 cf Appenota 0 to 10Cin. ?;rt ;0.
Since then, the i 11:enset has updated the perferin d ca analysis in accordance with Paragraph s
IV.A.1 of Appendix G to 10CFR Part 50 using criteria being developed by the American Society of Mechanical Engineers (ASME) to demonstrate margins of safety equivalent to those in the ASME Code.
l The NRC reviewed the cxprcastd ; senccra rc;;rding compliance with the requirements of Appenoix H to 10CFR. Part 50.
Section E 185 of the American Society for Testing and Haterials (ASTM) Code requires that the licensee take sample specimens from actual material used in f abricating the beltline of the reactor vessel. These surveillance materials shall incluoe one heat of base
} metal, one butt weld, and one weld "hcat affected zone." Yankee was placed j into operation before any requirements for surveillance programs were l promulgated. Nevertheless. Yankee implemented a Surveillance Program which
) helped form the basis for ASTM standards. The program was terminated in 1965 4
l due to structural f ailures of tne capsule holders.
In NUREG 0569. the staff
\\ concluded that sufficient information was generated to predict end of license l conditions. Howtyer, the pragram included only platt material. Ihc IICinici 4ct-inc iankcc Nucic;r Pe-ci Stetica tcraineted-the matcri;l ser+ci'lence prcgre; :n 1565 7-TherefGrc. th6-Y-tekte 'Lcicer Peuci StGtion hoc ne Letcriel
- uritillencc pregrea en July 25, lue0..hcr A;pcndix
- : te 10Cin re ; 50 LcceTc i
cffecti c.
Torthcr. tn: 3eripic;-45. cf+ttc; ct Ysakcc Re-c Ocforc 1:05 '.cic cenarisc;; cal 3 c+-tMri e-ee t eh R72\\49 2
ripR. 03 ' 92 14:20 YFraTE ATmIC ElICTRIC CO/BOLTON P.06 i
l l
The licensee for the Yankee Nuclear Power Station had used the methodology in Regulatory Guide 1.99. Revision 2. to predict the effects of neutron
] embrittlement. iicwcier. The staff raised concerns regarding the licensee's l application of the methodology, fe.nd th;t the actbedeiggy 1. ncs.ieter, ;sidt 1.^0. n;; ; ;i gr. 2.
.;; incer.cctij spplic t, tt; lic;as;;. The specific issues were (1) the irradiation temperature. (2) the chemistry composition of l reactor vessel weld material, and (3) the results of the material surveillance program.
The irradiation temperature at the Yankee Nuclear Power Station is between 454 xF and 520 xF. which is below the nominal irradiation temperature of 550 xF used in developing regulatory Guide 1.99. Revision 2.
A lower irradiation temperature increases the effect of neutron embrittlement. The regulatory guide indicates that for irradiation temperatures less tnan 525 xF.
enbrittlement effects should be considered to be greateh(than predicted by the j methods of the guide. The NRC disagreed with the adjustients thet -;rc made by the licensee w;rc in3ef ficient to account for this effect.
The liTitc0 results of the surveillance program from the Yankee Nuclear Power l Station indicated that the increase in the reference temperature for the plate l could exceect the rean-plus-two standard deviations as predicted by the procedures in Regulatory Guide 1.99, Revision 2.
The regulatory guide states that tne licensee should use credible surveillance data to predict the increase in reference temperature resulting frcm neutron irradiation. The staff imf emented Regulatory Guide 1.99. Revision 2. by issuing GL 88 11. In committing to GL 88-11. licensees have ccomitted to calculate radiation embrittlement in accordance with the procedures documented in RG 1.99 Revision 2.
To meet the limitations in Section 1.3 of the regulatory guide, the licensee shculd consider the affects on irradiation emorittlement during core critical operation with irradiation temperatures less than 525 xF.
Section 2 of the regulatory guide states that the licensees should consider the affects of the results from its surveillance capsules.
The Summer 1972 Addenda cf the 1971 Edition of Section !!I of the ASME Boiler and Pressure Ves.<el Code are the earliest ecde requirements for testing 1 materials to deternine their unirradiated reference temperature. 6tnee The l Yankee reactor vessel was constructed in 1958 tnd 1959 to ASME Section vill, ta : C0d; w;rlia tt;a thc !smacr 1072. it red not bc;n ; fficicnti, tcstcd l to dctcrainc ts Therefore, the unirradiated reference temperature could not l be in accordance with tc;;cc ts the reautrements of the Sumier 1972 Addenda.
l The licensee for the Yankee Nuclear Power Station used Branen Technical l Position MTEB 5 2 and extrece&e+ed the available test results to determine an R72\\49 3
^
- ~ ~
rFR. 03 ' 92 14:22 YR4TE AT0rGC ELECTRIC CCvDOLTON P. N unirradiated reference temperature. Th et;f f dcterr.b. ; Get th; ;
i.muu
. w
..yw
...... ov. wwn.s..
6..s.
The chemical composition of the Yankee reactor vessel welds is unknown.
The material's sensitivity to neutron embrittlement depends on its chemical content.
The licencee assumed that the chemistry of its welds was equivalent to that of the BR-3 reactor vessel in Hol, Belgium.
l cc, i d 7,ct i d;.r ti f
. n,
. u....
f The heat number of the wire used to fabricate the Yankee
{ welds was not available.
The licensee was assuming a chemical composition that was not based on its plant specific information, since the chemical composition, in particular, the amount of copper, depends upon the heat number of the weld wire, l These factors prompted the staff to question how fir.; u.Li the licensee for the Yankee Nuclear Power Station had not considered plant spectiic inferination l in assessing compliance with 10CFR50.61.
Since there is no When l plant specific information, the licensee used bounding values of weld l chemistry established by the staf f.
Using these bounding values, the staff
[ estimated dcter ciried that ia cei. sic;ric, the Yankee reactor vessel could mey
( have exceeded the screening criteria in 10CFR50.61. Therefore :inc; then, the licensee has performed a probabilistic fracture mechanics analysis in accor dance with 10CFR50.61(b) (4) and the staff is continuing its review.
Upon conducting the Yankee Nuclear Power Station review, the staff became l concerned about other licensee *s conpliance tht this mej oct bc ;. '.;r/.etid cesc isgerdhs ;.;,aliencc with 10CFR50.60 and 10CFR50.61 anc fulfillment of commitments made in response to GL 8911. Thus, the staff is issuing this generic letter to obtain information to assess compliance with these regulations and fulfillment of commitments. The staff is continuing to pursue this concern with the Yankee Atomic Electric Company.
Therefore, the Yankee Atomic Electric Company need not respond to this generic letter.
Required Information Portions of the following information requested are not applicable to all addressees.
The responses provided should, in these cases, indicate that the requested inforttation is rot applicable and why it is not applicable.
9202250115 Y R72\\49 4
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