Reactor Vessel Structural Integrity, 10 CFR 50.54(f) (Generic Letter 92-01, Revision 1)

TO:       ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FOR 
          NUCLEAR POWER PLANTS (EXCEPT YANKEE ATOMIC ELECTRIC COMPANY, 
          LICENSEE FOR THE YANKEE NUCLEAR POWER STATION)

SUBJECT:  REACTOR VESSEL STRUCTURAL INTEGRITY, 10 CFR 50.54(f)
          (GENERIC LETTER 92-01, REVISION 1)


This letter replaces Generic Letter 92-01 dated February 28, 1992.  The 
background information concerning NRC's assessment of embrittlement in the 
Yankee Nuclear Power Station 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. 

The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter 
to obtain information needed to assess compliance with requirements and 
commitments 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) 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 is necessary pursuant to 10 
CFR 50.12.  In 10 CFR 50.61 the NRC also provided fracture toughness 
requirements for protecting pressurized 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 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 10 CFR Part 
50, Appendix G.  The 10 CFR 50.60 and 10 CFR 50.61 requirements and GL 88-11 
are in the overall regulatory program to maintain the structural integrity 
of the reactor vessel.  

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 commitments made in response to GL 88-11.  Enclosure 1 is a 
discussion of the applicable regulatory requirements.  The NRC is requiring 
information on compliance under the provisions of 10 CFR 50.54(f).

                                    - 2 -



Assessment of Embrittlement for the Yankee Nuclear Power 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 
reactor vessel.  The staff found that the licensee for the Yankee Nuclear 
Power Station might not be in compliance with 10 CFR 50.60 and 10 CFR 50.61.  

The staff found that 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.  However, the 
licensee for the Yankee Nuclear Power Station had not performed the actions 
required in Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.  Since 
then, the licensee has performed an analysis in accordance with Paragraph 
IV.A.1 of Appendix G to 10 CFR 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.

The NRC expressed a concern regarding compliance 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 surveillance 
program in 1965.  Therefore, the Yankee Nuclear Power Station had no 
material surveillance program on July 26, 1983, when Appendix H to 10 CFR 
Part 50 became effective.  Further, the samples irradiated at Yankee Rowe 
before 1965 were comprised only of base metal.

The licensee for the Yankee Nuclear Power Station had used the methodology 
in draft Regulatory Guide 1.99, Revision 2, to predict the effects of 
neutron embrittlement.  The staff raised concerns regarding the licensee's 
application of the methodology. The specific issues were (1) the irradiation 
temperature, (2) the chemistry composition of reactor vessel material, and 
(3) the results of the material surveillance program.

The irradiation temperature at the 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 temperature increases the effect of neutron embrittlement.  The 
regulatory guide indicates that for irradiation temperatures less than 
525 �F, embrittlement effects should be considered to be greater than 
predicted by the methods of the guide.  Adjustments that were made by the 
licensee were insufficient to account for this effect.  

The 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 use credible surveillance data to predict the increase in 
reference temperature resulting from neutron irradiation.  

.
                                    - 3 -



The staff implemented RG 1.99, Revision 2, by issuing GL 88-11.  In 
committing to GL 88-11, licensees have committed 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 Summer 1972 Addenda of the 1971 Edition of Section III of the ASME 
Boiler and Pressure Vessel Code are the earliest code requirements for 
testing materials to determine their unirradiated reference temperature.  
The Yankee reactor vessel was constructed in 1959 to ASME Code, Section 
VIII. Therefore, the unirradiated reference temperature could not be 
established in accordance with the requirements of the Summer 1972 Addenda.  
The licensee for the Yankee Nuclear Power Station extrapolated the available 
test results to determine an unirradiated reference temperature.  The staff 
determined that the licensee's extrapolation was not conservative.  

The chemical composition of the Yankee reactor vessel welds is unknown.  The 
material's sensitivity to neutron embrittlement depends on its chemical 
content.  The licensee assumed that the chemistry of its welds was 
equivalent to that of the BR-3 reactor vessel in Mol, Belgium.  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.

These factors prompted the staff to find that the licensee for the Yankee 
Nuclear Power Station had not fully considered plant-specific information in 
assessing compliance with 10 CFR 50.61.  When plant-specific information is 
considered, the Yankee reactor vessel may have exceeded the screening 
criteria in 10 CFR 50.61.  

Upon conducting the Yankee Nuclear Power Station review, the staff became 
concerned about other licensee's 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.  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 information is not applicable and why it is not applicable. 

.
                                    - 4 -


1.   Certain addressees are requested to provide the following information 
     regarding Appendix H to CFR Part 50:

          Addressees who do not have a surveillance program meeting ASTM E 
          185-73, -79, or -82 and who do not have an integrated surveillance 
          program approved by the NRC (see Enclosure 2), are requested to 
          describe actions taken or to be taken to ensure compliance with 
          Appendix H to 10 CFR Part 50.  Addressees who plan to revise the 
          surveillance program to meet Appendix H to 10 CFR Part 50 are 
          requested to indicate when the revised program will be submitted 
          to the NRC staff for review.  If the surveillance program is not 
          to be revised to meet Appendix H to 10 CFR Part 50, addressees are 
          requested to indicate when they plan to request an exemption from 
          Appendix H to 10 CFR Part 50 under 10 CFR 50.60(b). 

2.   Certain addressees are requested to provide the following 
     information regarding Appendix G to 10 CFR Part 50:

     a.   Addressees of plants for which the Charpy upper shelf energy is 
          predicted to be less than 50 foot-pounds at the end of their 
          licenses using the guidance in Paragraphs C.1.2 or C.2.2 in 
          Regulatory Guide 1.99, Revision 2, are requested to provide to the 
          NRC the Charpy upper shelf energy predicted for December 16, 1991, 
          and for the end of their current license for the limiting beltline 
          weld and the plate or forging and are requested to describe the 
          actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G 
          to 10 CFR Part 50.  

     b.   Addressees whose reactor vessels were constructed to an ASME Code 
          earlier than the Summer 1972 Addenda of the 1971 Edition are 
          requested to describe the consideration given to the following 
          material properties in their evaluations performed pursuant to 10 
          CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:

          (1)  the results from all Charpy and drop weight tests for all 
               unirradiated beltline materials, the unirradiated reference 
               temperature for each beltline material, and the method of 
               determining the unirradiated reference temperature from the 
               Charpy and drop weight test;

          (2)  the heat treatment received by all beltline and surveillance 
               materials;

          (3)  the heat number for each beltline plate or forging and the 
               heat number of wire and flux lot number used to fabricate 
               each beltline weld;

.
                                    - 5 -


          (4)  the heat number for each surveillance plate or forging and 
               the heat number of wire and flux lot number used to fabricate 
               the surveillance weld;

          (5)  the chemical composition, in particular the weight in percent 
               of copper, nickel, phosphorous, and sulfur for each beltline 
               and surveillance material; and 

          (6)  the heat number of the wire used for determining the weld 
               metal chemical composition if different than Item (3) above.

3.   Addressees are requested to provide the following information regarding 
     commitments made to respond to GL 88-11: 
     
     a.   How the embrittlement effects of operating at an irradiation 
          temperature (cold leg or recirculation suction temperature) below 
          525 �F were considered.  In particular licensees are requested to 
          describe consideration given to determining the effect of lower 
          irradiation temperature on the reference temperature and on the 
          Charpy upper shelf energy.

     b.   How their surveillance results on the predicted amount of 
          embrittlement were considered. 
          
     c.   If a measured increase in reference temperature exceeds the 
          mean-plus-two standard deviations predicted by Regulatory Guide 
          1.99, Revision 2, or if a measured decrease in Charpy upper shelf 
          energy exceeds the value predicted using the guidance in Paragraph 
          C.1.2 in Regulatory Guide 1.99, Revision 2, the licensee is 
          requested to report the information and describe the effect of the 
          surveillance results on the adjusted reference temperature and 
          Charpy upper shelf energy for each beltline material as predicted 
          for December 16, 1991, and for the end of its current license.

Reporting Requirements

Pursuant to Section 182a of the Atomic Energy Act of 1954, as amended, and 
10 CFR 50.54(f), each addressee shall submit a letter within 120 days of the 
date of this generic letter providing the information described under 
"Required Information."  The letter shall be addressed to the U.S. Nuclear 
Regulatory Commission, ATTN:  Document Control Desk, Washington, DC 20555, 
under oath or affirmation.  A copy shall also be submitted to the 
appropriate Regional Administrator.  This generic letter requests 
information that will enable the NRC to verify that the licensee is 
complying with its current licensing basis regarding reactor vessel fracture 
toughness and material surveillance for the reactor coolant pressure 
boundary.  Accordingly, an evaluation justifying this information request is 
not necessary under 10 CFR 50.54(f).
.
                                    - 6 -


Backfit Discussion

This generic letter requests information that will enable the NRC staff to 
determine whether licensees are complying with their prior commitments and 
any license conditions regarding 10 CFR 50.60, 10 CFR 50.61, and GL 88-11.  
The staff is not establishing a new position for such compliance in this 
generic letter.  The staff is requesting information to verify that the 
licensee is complying with its previously established commitments and is not 
establishing any new position.  Therefore, this generic letter does not 
constitute a backfit and no documented evaluation or backfit analysis need 
be prepared.  

Request for Voluntary Submittal of Impact Data

This request is covered by Office of Management and Budget Clearance Number 
3150-0011, which expires May 31, 1994.  The estimated average number of 
burden hours is 200 person hours for each addressee's response, including 
the time required to assess the requirements, search data sources, gather 
and analyze the data, and prepare the required letters.  This estimated 
average number of burden hours pertains only to the identified 
response-related matters and does not include the time to implement the 
actions required by the regulations.  Comments on the accuracy of this 
estimate and suggestions to reduce the burden may be directed to Ronald 
Minsk, Office of Information and Regulatory Affairs (3150-0011), NEOB-3019, 
Office of Management and Budget, Washington, DC 20503, and to the U.S. 
Nuclear Regulatory Commission, Information and Records Management Branch, 
Division of Information Support Services, Office of Information and 
Resources Management, Washington, DC 20555.

Although no specific request or requirement is intended, the following 
information would assist the NRC in evaluating the cost of complying with 
this generic letter:

(1)  the licensee staff's time and costs to perform requested inspections, 
     corrective actions, and associated testing;

(2)  the licensee staff's time and costs to prepare the requested reports 
     and documentation;

(3)  the additional short-term costs incurred to address the inspection 
     findings such as the costs of the corrective actions or the costs of 
     down time; and 
     
(4)  an estimate of the additional long-term costs that will be incurred as 
     a result of implementing commitments such as the estimated costs of 
     conducting future inspections or increased maintenance.
.
      
                                    - 7 -



If you have any questions about this matter, please contact one of the NRC 
technical contacts or the lead project manager listed below.

                                   Sincerely,



                                   James G. Partlow 
                                   Associate Director for Projects
                                   Office of Nuclear Reactor Regulation

Enclosures:
1.  Applicable Regulatory Requirements
2.  Plants with Integrated Programs
3.  List of Recently Issued Generic Letters 

Technical Contacts:
Barry J. Elliot, NRR
(301) 504-2709

Keith R. Wichman, NRR
(301) 504-2757

Lead Project Manager:
Daniel G. McDonald, NRR
(301) 504-1408 
.

                                                          Enclosure 1

                   Regulatory Requirements Applicable to 

                    Reactor Vessel Structural Integrity 

10 CFR 50.60

Pursuant to 10 CFR 50.60, all light water nuclear power reactors must meet 
the fracture toughness and material surveillance program requirements for 
the reactor coolant pressure boundary set forth in Appendices G and H to 10 
CFR Part 50.

The fracture toughness of the reactor coolant pressure boundary required by 
10 CFR 50.60 is necessary to provide adequate margins of safety during any 
condition of normal plant operation, including anticipated operational 
occurrences and system hydrostatic tests.  The material surveillance program 
required by 10 CFR 50.60 monitors changes in the fracture toughness 
properties of ferritic materials in the reactor vessel beltline region of 
light water nuclear power reactors resulting from exposure of these 
materials to neutron irradiation and the thermal environment.  Under the 
program, fracture toughness test data are obtained from material specimens 
exposed in surveillance capsules, which are withdrawn periodically from the 
reactor vessel.

Appendix G to 10 CFR Part 50 requires that the reactor vessel beltline 
materials must have Charpy upper shelf energy of no less than 50 ft-lb 
throughout the life of the vessel.  Otherwise, licensees are required to 
provide demonstration of equivalent margins of safety in accordance with 
Paragraph IV.A.1 of Appendix G to 10 CFR Part 50 or perform actions in 
accordance with Paragraph V.C of Appendix G to 10 CFR Part 50.

Appendix H to 10 CFR Part 50 requires the surveillance program to meet the 
American Society for Testing and Materials (ASTM) Standard E 185, "Standard 
Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear 
Power Reactor Vessels."  Further, Appendix H to 10 CFR Part 50 specifies the 
applicable edition of ASTM E 185.  Appendix H to 10 CFR Part 50, as amended 
on July 26, 1983, requires that the part of the surveillance program 
conducted before the first capsule is withdrawn must meet the requirements 
of the 1973, the 1979, or the 1982 edition of ASTM E 185 that is current on 
the issue date of the American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel Code under which the reactor vessel was purchased.  The 
licensee may also use later editions of ASTM E 185 which have been endorsed 
by the NRC.  The test procedures and reporting requirements for each capsule 
withdrawal after July 26, 1983 must meet the requirements of the 1982 
edition of ASTM E 185 to the extent practical for the configuration of the 
specimens in the capsule.  The licensee may use either the 1973, the 1979, 
or the 1982 edition of ASTM E 185 for each capsule withdrawal before 
July 26, 1983.
.
                                    - 2 -



Licensees, especially those with reactor vessels purchased before ASTM 
issued the 1973 edition of ASTM E 185, may have surveillance programs that 
do not meet the requirements of Appendix H to 10 CFR Part 50 but may have 
alternative surveillance programs.  The licensee may use these alternative 
surveillance programs in accordance with 10 CFR 50.60(b) if the licensee has 
been granted an exemption by the Commission under 10 CFR 50.12.

The licensee must monitor the test results from the material surveillance 
program.  According to Paragraph III.C of Appendix H to 10 CFR Part 50, the 
results of the surveillance program may indicate that a technical 
specifications change is required, either in the pressure-temperature limits 
or in the operating procedures required to meet the limits.

10 CFR 50.61

Pursuant to 10 CFR 50.61, there are fracture toughness requirements for 
protection against pressurized thermal shock events for pressurized water 
reactors.  Licensees are required to perform an assessment of the projected 
values of reference temperature.  If the projected reference temperature 
exceeds the screening criteria established in 10 CFR 50.61, licensees are 
required to submit an analysis and schedule for such flux reduction programs 
as are reasonably practicable to avoid exceeding the screening criteria.  If 
no reasonably practicable flux reduction program will avoid exceeding the 
screening criteria, licensees shall submit a safety analysis to determine 
what actions are necessary to prevent potential failure of the reactor 
vessel if continued operation beyond the screening criteria is allowed.  In 
10 CFR 50.61(b)(1), as amended effective June 14, 1991 (56 Fed Reg 22300 et. 
seq., May 15, 1991), licensees are required to submit their assessment by 
December 16, 1991, if the projected reference temperature will exceed the 
screening criteria before the expiration of the operating license.

Plant-specific information is required to be considered in assessing the 
level of neutron embrittlement as specified in 10 CFR 50.61(b)(3).  This 
information includes but is not limited to the reactor vessel operating 
temperature and surveillance results.

Prediction of Irradiation Embrittlement

Paragraph V.A of Appendix G to 10 CFR Part 50 requires the prediction of the 
effects of neutron irradiation on reactor vessel materials.  The extent of 
neutron embrittlement depends on the material properties, thermal 
environment, and results of the material surveillance program.  In Generic 
Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel 
Materials and its Impact on Plant Operations," the staff stated that it will 
use the guidance in Regulatory Guide 1.99, Revision 2, "Radiation 
Embrittlement of Reactor Vessel Materials," in estimating the embrittlement 
of the materials in the reactor vessel beltline.  All licensees and 
permittees have responded to Generic Letter 88-11 committing to use the 
methodology in Regulatory Guide 1.99, 

.
                                    - 3 -



Revision 2, in predicting the effects of neutron irradiation as required by 
Paragraph V.A of 10 CFR Part 50, Appendix G.  The methodology in Regulatory 
Guide 1.99, Revision 2, is also the basis in 10 CFR 50.61 in projecting the 
reference temperature.

                                                       Enclosure 2


Plants With Integrated Surveillance Programs Approved By The NRC


                         Oconee Units 1, 2, and 3
                         Arkansas Nuclear One Unit 1
                         Rancho Seco 
                         Three Mile Island Unit 1 
                         Davis-Besse 
                         Ginna 
                         Point Beach Units 1 and 2 
                         Surry Units 1 and 2 
                         Turkey Point Units 3 and 4 
                         Zion Units 1 and 2