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Home       > Electronic Reading       Room       > Document       Collections       > Generic       Communications >       Generic       Letters >       1997       > GL97001 
				UNITED STATES                         NUCLEAR REGULATORY COMMISSION                      OFFICE OF NUCLEAR REACTOR REGULATION                           WASHINGTON, D.C. 20555-0001                                  April 1, 1997NRC GENERIC LETTER 97-01:  DEGRADATION OF CONTROL ROD DRIVE MECHANISM                                     NOZZLE AND OTHER VESSEL CLOSURE HEAD                                           PENETRATIONS AddresseesAll holders of operating licenses for pressurized water reactors (PWRs),except those who have permanently ceased operations and have certified thatfuel has been permanently removed from the reactor vessel. PurposeThe U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to(1) request addressees to describe their program for ensuring the timelyinspection of PWR control rod drive mechanism (CRDM) and other vessel closurehead penetrations and (2) require that all addressees provide to the NRC awritten response to the requested information.  The information requested isneeded by the NRC staff to verify compliance with 10 CFR 50.55a and 10 CFRPart 50, Appendix A, GDC 14, and to determine whether an augmented inspectionprogram, pursuant to 10 CFR 50.55a(g)(6)(ii), is required.BackgroundPrimary Water Stress Corrosion Cracking of Vessel Closure Head PenetrationsMost PWRs have Alloy 600 CRDM nozzle and other vessel head closurepenetrations (VHPs) that extend above the reactor pressure vessel head.  Thestainless steel housing of the CRDM is screwed and seal-welded onto the top ofthe nozzle penetration, as shown in  Figure 1.  (Figure 1 is for illustrativepurposes only and is not intended to be indicative of every nuclear steamsupply system (NSSS) vendor's CRDM design.)  The weld between the nozzle topand bottom pieces is a dissimilar metal weld, which is also called abimetallic weld.  The nozzles protrude below the vessel head, thus exposingthe inside surface of the nozzles to reactor coolant.  The CRDM nozzle andother VHPs are basically the same for all PWRs worldwide, which use a U.S.design (except in Germany and Russia).  The areas of interest for potentialcracking are the weld between the nozzle and reactor vessel head, and theportion of the nozzle inside the reactor vessel head above the nozzle-to-vessel weld.Generally, there are 36 to 78 nozzles distributed over the low-alloy steelhead.  The vessel head is semi-spherical and the head penetrations arevertical so that the CRDM nozzle and other VHPs are not perpendicular to thevessel surface except at the center.  The uphill side (toward the center ofthe head) is called the 180-degree location and the downhill side (toward theouter periphery of the head) is called the 0-degree location.  Most nozzleshave a thermal sleeve with a conical guide at the bottom end and a small gap(3- to 4-mm) [0.12 to 0.16 in.] between the nozzle and the sleeve.9703260336.                                                                                                                                                                                                            GL 97-01                                                                                                                                                                                                            April 1, 1997                                                                                                                                                                                                            Page 2 of 10Beginning in 1986, leaks have been reported in several Alloy 600 pressurizerinstrument nozzles at both domestic and foreign reactors from severaldifferent NSSS vendors.  The NRC staff identified primary water stresscorrosion cracking (PWSCC) as an emerging technical issue to the Commission in1989, after cracking was noted in Alloy 600 pressurizer heater sleevepenetrations at a domestic PWR facility.  The NRC staff reviewed the safetysignificance of the cracking that occurred, as well as the repair andreplacement activities at the affected facilities.  The NRC staff determinedthat the cracking was not of immediate safety significance because the crackswere axial, had a low growth rate, were in a material with an extremely highflaw tolerance (high fracture toughness) and, accordingly, were unlikely topropagate very far.  These factors also demonstrated that any cracking wouldresult in detectable leakage and the opportunity to take corrective actionbefore a penetration would fail.  Further, with the exception of the leakfound at Bugey 3 during hydrostatic testing, the NRC staff is not aware of anyfailure of an Alloy 600 vessel closure head penetration during plantoperation.  The NRC staff issued Information Notice (IN) 90-10, "Primary WaterStress Corrosion Cracking (PWSCC) of Inconel 600," dated February 23, 1990, toinform the nuclear industry of the issue.In September 1991, cracks were found in an Alloy 600 VHP in the reactor headat Bugey 3, a French PWR.  Examinations in PWRs in France, Belgium, Sweden,Switzerland, Spain, and Japan were performed, and additional VHPs with axialcracks were detected in several European plants.  About 2 percent of the VHPsexamined to date contain short, axial cracks.  Close examination of the VHPthat leaked at Bugey 3 revealed very minor incipient secondary circumferentialcracking of the VHP.  European and Japanese utilities have taken steps todetect and mitigate the PWSCC damage and to detect the leakage at an earlystage.  European and Japanese utilities have inspected most of the CRDMnozzles and repaired the nozzles or replaced the vessel heads as appropriate. In Japan, the three most susceptible vessel heads are being replaced, eventhough no cracks were found in the nozzles of these heads.  In France,?lecricit‚ de France (EdF) is planning on replacing all vessel heads as apreventative measure.  Inservice inspection of the upper head is now requiredin Sweden.   Removable insulation on the vessel head and leakage monitoringsystems are installed at French and Swedish plants for early detection ofleakage.An action plan was implemented by the NRC staff in 1991 to address PWSCC ofAlloy 600 VHPs at all U.S. PWRs.  As explained more fully below, this actionplan included a review of the safety assessments by the PWR Owners Groups, thedevelopment of VHP mock-ups by the Electric Power Research Institute (EPRI),the qualification of inspectors on the VHP mock-ups by EPRI, the review ofproposed generic acceptance criteria from the Nuclear Utility Management andResource Council (NUMARC) [now the Nuclear Energy Institute (NEI)], and VHPinspections.  As part of this action plan, the NRC staff met with the Westinghouse Owners Group (WOG) on January 7, 1992, the Combustion EngineeringOwners Group (CEOG) on March 25, 1992, and the Babcock & Wilcox Owners Group(B&WOG) on May 12, 1992, to discuss their respective programs forinvestigating PWSCC of Alloy 600 and to assess the possibility of cracking ofVHPs in their respective plants since all of the plants have Alloy 600 VHPs. Subsequently, the NRC staff asked NUMARC to coordinate future industry actionsbecause the issue was applicable to all PWRs.  Meetings .                                                                                                                                                                                                            GL 97-01                                                                                                                                                                                                            April 1, 1997                                                                                                                                                                                                            Page 3 of 10were held with NUMARC/NEI and the PWR Owner's Groups on the issue on August 18and November 20, 1992, March 3, 1993, December 1, 1994, and August 24, 1995. Summaries of these meetings are available in the Commission's Public DocumentRoom, 2120 L Street, N.W., Washington, D.C. 20555.Each of the PWR Owners Groups submitted safety assessments, dated February1993, through NUMARC to the NRC on this issue.  After reviewing the industry'ssafety assessments and examining the overseas inspection findings, the NRCstaff concluded in a safety evaluation dated November 19, 1993, that VHPcracking was not an immediate safety concern.  The bases for this conclusionwere that if PWSCC occurred at VHPs (1) the cracks would be predominatelyaxial in orientation, (2) the cracks would result in detectable leakage beforecatastrophic failure, and (3) the leakage would be detected during visualexaminations performed as part of surveillance walkdown inspections beforesignificant damage to the reactor vessel closure head would occur.  Inaddition, the NRC staff had concerns related to unnecessary occupationalradiation exposures associated with eddy current or other forms ofnondestructive examinations (NDEs), if performed manually.  Field experiencein foreign countries has shown that occupational radiation exposures can besignificantly reduced by using remotely controlled or automatic equipment toconduct the inspections.In 1993, the nuclear industry developed remotely operated inservice inspectionequipment and repair tools that reduced radiation exposure.  Techniques andprocedures developed by two vendors were successfully demonstrated in a blindqualification protocol developed and administered by the EPRI NDE Center.  Inthe demonstrations, examinations by rotating and saber eddy current andultrasonics showed a high probability of detection of the flaws which werealso sized within reasonable uncertainty bounds.  The qualification testingalso demonstrated that personnel qualified through the EPRI program canreliably detect PWSCC in CRDM nozzles.Intergranular Attack of CRDM Penetration Nozzle at ZoritaIn 1994, circumferential intergranular attack (IGA) associated with the weldbetween the inner surface of the reactor closure head and the CRDM penetration(usually referred to as the J-grove weld) in one of the CRDM penetrations wasdiscovered at Zorita, a Spanish reactor.  This IGA is a different degradationmechanism than the PWSCC described above.  It is believed to have resultedfrom the combination of ion exchange resin bead intrusions, which resulted inhigh concentrations of sulfates.  Zorita has 37 CRDM penetrations, of which 20are active penetrations and 17 are spare penetrations.  Sixteen of the 17spare penetrations showed stress corrosion cracking and IGA.  The cracks wereboth axial and circumferential.  Four of the active CRDM penetrations hadsignificant cracking with axial and circumferential cracks.  Two cation resiningress events occurred at Zorita.  In August 1980, 40 liters [10.57 U.S.gallons] of cation resin entered the reactor coolant system (RCS).  InSeptember 1981, a mixed bed demineralizer screen failed and between 200 to 320liters [52.83 to 84.54 U.S. gallons] of resin entered the RCS.  The coolantconductivity remained high for at least 4 months after the ingress.  Theincrease in conductivity was attributed to locally high .                                                                                                                                                                                                            GL 97-01                                                                                                                                                                                                            April 1, 1997                                                                                                                                                                                                            Page 4 of 10concentrations of sulfates.  Sulfates were found around the crack areas and onthe fracture surfaces.  It is important to note that sulfate cracking canoccur in regions that are not subject to significant applied or residualstresses.The NRC staff issued IN 96-11, "Ingress of Demineralizer Resins IncreasesPotential for Stress Corrosion Cracking of Control Rod Drive MechanismPenetrations," dated  February 14, 1996, to alert addressees to the increasedlikelihood of sulfate-driven stress corrosion cracking of PWR CRDMs and otherVHPs if demineralizer resins contaminate the RCS. Westinghouse notified the WOG plants, the B&WOG plants, and the CEOG plants ofthe Zorita incident by issuing NSAL-94-028.  Westinghouse reported that noother plant had been found worldwide that had experienced cracking similar tothat at the Zorita plant.  Westinghouse further reported that U.S. plantsmonitor RCS conductivity on a routine basis, follow the EPRI guidelines onprimary water chemistry, and monitor for sulfate three times a week. Westinghouse concluded that no immediate safety issue is involved and that theconclusions in its CRDM safety evaluation remain valid.  Westinghousesuggested that U.S. PWR plants review their RCS chemistry and other operatingrecords pertaining to sulfur ingress events.  The results of this review havenot been reported to the NRC staff, and the NRC staff does not have sufficientinformation to ascertain whether any significant primary system resin beadintrusions have occurred at any U.S. PWR.The first U.S. inspection of VHPs took place in the spring of 1994 at thePoint Beach Nuclear Generating Station, and no indications were detected inany of its 49 CRDM penetrations.  The eddy current inspection at the OconeeNuclear Generating Station in the fall of 1994 revealed 20 indications in onepenetration.  Ultrasonic testing (UT) did not reveal the depth of theseindications because they were shallow.  UT cannot accurately size defects thatare less than one mil deep (0.03 mm).  These indications may be associatedwith the original fabrication and may not grow; however, they will bereexamined during the next refueling outage.  A limited examination of eightin-core instrumentation penetrations conducted at the Palisades plant found nocracking.  An examination of the CRDM penetrations at the  D. C. Cook plant inthe fall of 1994 revealed three clustered indications in one penetration.  Theindications were 46 mm [1.81 in.], 16 mm [0.63 in.], and 6 to 8 mm [0.24 to0.31 in.] in length, and the deepest flaw was 6.8 mm [0.27 in.] deep.  The tipof the 46-mm [1.81 in.] flaw was just below the J-groove weld.Virginia Electric and Power Company inspected North Anna Unit 1 during itsspring 1996 refueling outage.  Some high-stress areas (e.g., upper and lowerhillsides) were examined on each outer ring CRDM penetrations and noindications were observed using eddy current testing.The NRC staff was informed during a meeting on August 24, 1995, thatWestinghouse had developed a susceptibility model for VHPs based on a numberof factors, including operating temperature, years of power operation, methodof fabrication of the VHP, microstructure of .                                                                                                                                                                                                            GL 97-01                                                                                                                                                                                                            April 1, 1997                                                                                                                                                                                                            Page 5 of 10the VHP, and the location of the VHP on the head.  Each time a plant's VHPsare inspected, the inspection results are incorporated into the model.  Alldomestic Westinghouse PWRs have been modeled and the ranking has been given toeach licensee.  In addition, the NRC staff was informed that FramatomeTechnologies, Inc. [FTI, formerly Babcock & Wilcox (B&W)], also developed asusceptibility model for CRDM penetration nozzles and other VHPs in B&Wreactor vessel designs.  All domestic B&W PWRs have been modeled and theranking has been given to each B&W licensee.  The NRC staff was furtherinformed that Combustion Engineering (CE) had performed an initialsusceptibility assessment for the CE PWRs.  At present, none of the PWR OwnersGroups (i.e., WOG, B&WOG, or CEOG) has submitted its models and assessments tothe NRC staff for review.By letter dated March 5, 1996, NEI submitted a white paper entitled "Alloy 600RPV Head Penetration Primary Stress Corrosion Cracking," which reviews thesignificance of PWSCC in PWR VHPs and describes how the industry is managingthe issue.  The program outlined in the NEI white paper is based on theassumption that the issue is primarily an economic rather than a safety issue,and describes an economic decision tool to be used by PWR licensees toevaluate the probability of a VHP developing a crack or a through-wall leakduring a plant's lifetime.  This information would then be used by a PWRlicensee to evaluate the need to conduct a VHP inspection at their plant.  TheNRC staff informed NEI in the several meetings listed above that it did notagree with NEI that the issue was primarily economic.DiscussionThe results of domestic VHP inspections are consistent with the February 1993analyses by the PWR Owners Groups, the NRC staff safety evaluation reportdated November 19, 1993, and the PWSCC found in the CRDMs in Europeanreactors.  On the basis of the results of the first five inspections of U.S.PWRs, the PWR Owner's Groups' analyses, and the European experience, the NRCstaff has determined that it is probable that VHPs at other plants containsimilar axial cracks.  Further, if any significant resin intrusions have occurred at U.S. PWRs such as occurred at Zorita, residual stresses aresufficient to cause circumferential intergranular stress corrosion cracking(IGSCC).After considering this information, the NRC staff has concluded that VHPcracking does not pose an immediate or near term safety concern.  Further, theNRC staff recognizes that the scope and timing of inspections may vary fordifferent plants depending on their individual susceptibility to this form ofdegradation.  In the long term, however, degradation of the CRDM and otherVHPs is an important safety consideration that warrants further evaluation. The vessel closure head provides the vital function of maintaining reactorpressure boundary.  Cracking in the VHPs has occurred and is expected tocontinue to occur as plants age.  The NRC staff considers cracking of VHPs tobe a safety concern for the long term based on the possibility of (1) exceeding the American Society of Mechanical Engineers(ASME) Code for margins if the cracks are sufficiently deep and continue topropagate during subsequent operating cycles, and (2) eliminating a layer ofdefense in depth for plant safety.  Therefore, .                                                                                                                                                                                                            GL 97-01                                                                                                                                                                                                            April 1, 1997                                                                                                                                                                                                            Page 6 of 10to verify that the margins required by the ASME Code, as specified in Section50.55a of Title 10 of the Code of Federal Regulations (10 CFR 50.55a) are met,that the guidance of General Design Criterion 14 of Appendix A to 10 CFRPart 50 (10 CFR Part 50, Appendix A, GDC 14) is continued to be satisfied, andto ensure that the safety significance of VHP cracking remains low, the NRCstaff continues to believe that an integrated, long-term program, whichincludes periodic inspections and monitoring of VHPs, is necessary.  This wasthe conclusion of the staff's November 19, 1993, safety evaluation, whichstated, in part, "...the staff recommends that you consider enhanced leakagedetection by visually examining the reactor vessel head until eitherinspections have been completed showing absence of cracking or on-line leakagedetection is installed in the head area ... nondestructive examinations shouldbe performed to ensure there is no unexpected cracking in domestic PWRs. These examinations do not have to be conducted immediately ... As thesurveillance walkdowns proposed by NUMARC are not intended for detecting smallleaks, it is conceivable that some affected PWRs could potentially operatewith small undetected leakage at CRDM/CEDM penetrations.  In this regard, thestaff believes that it is prudent for NUMARC to consider the implementation ofan enhanced leakage detection method for detecting small leaks during plantoperation."  In addition, the NRC staff finds that the requested informationis also needed to determine if the imposition of an augmented inspectionprogram, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain publichealth and safety.The NRC staff recognizes that individual PWR licensees may wish to determinetheir inspection activities based on an integrated industry inspection program(i.e., B&WOG, CEOG, WOG, or some subset thereof), to take advantage ofinspection results from other plants that have similar susceptibilities.  TheNRC staff does not discourage such group actions but notes that such anintegrated industry inspection program must have a well-founded technicalbasis that justifies the relationship between the plants and the plannedimplementation schedule.Requested InformationThe information requested in item 1 is needed by the NRC staff to verifycompliance with 10 CFR 50.55a and 10 CFR Part 50, Appendix A, GDC 14, and todetermine whether an augmented inspection program of the weld between thepenetration nozzle and reactor vessel head as well as the portion of thenozzle above the weld is required, pursuant to 10 CFR 50.55a(g)(6)(ii), whilethe information requested in item 2 relates to the occurrence of resin beadintrusion in PWRs, such as occurred at Zorita.Within 120 days of the date of this generic letter, each addressee isrequested to provide a written report that includes the following informationfor its facility:.                                                                                                          GL 97-01                                                                                                          April 1, 1997                                                                                                                                                                                                           Page 7 of 101.           Regarding inspection activities:              1.1              A description of all inspections of CRDM nozzle and other VHPs                              performed to the date of this generic letter, including the results                              of these inspections.             1.2              If a plan has been developed to periodically inspect the CRDM nozzle                              and other VHPs:                              a.           Provide the schedule for first, and subsequent, inspections of                                           the CRDM nozzle and other VHPs, including the technical basis for                                           this schedule.                              b.           Provide the scope for the CRDM nozzle and other VHP inspections,                                           including the total number of penetrations (and how many will be                                           inspected), which penetrations have thermal sleeves, which are                                           spares, and which are instrument or other penetrations.             1.3              If a plan has not been developed to periodically inspect the CRDM                              nozzle and other VHPs, provide the analysis that supports why no                              augmented inspection is necessary.              1.4              In light of the degradation of CRDM nozzle and other VHPs described                              above, provide the analysis that supports the selected course of                              action as listed in either 1.2 or 1.3, above.  In particular, provide                              a description of all relevant data and/or tests used to develop crack                              initiation and crack growth models, the methods and data used to                              validate these models, the plant-specific inputs to these models, and                              how these models substantiate the susceptibility evaluation.  Also,                              if an integrated industry inspection program is being relied on,                              provide a detailed description of this program.2.           Provide a description of any resin bead intrusions, as described in IN 96-             11, that have exceeded the current EPRI PWR Primary Water Chemistry             Guidelines recommendations for primary water sulfate levels, including the             following information:             2.1              Were the intrusions cation, anion, or mixed bed?             2.2              What were the durations of these intrusions?             2.3              Does the plant's RCS water chemistry Technical Specifications follow                              the EPRI guidelines? .                                                                                                          GL 97-01                                                                                                          April 1, 1997                                                                                                                                                                                                           Page 8 of 10             2.4              Identify any RCS chemistry excursions that exceed the plant                              administrative limits for the following species:  sulfates, chlorides                              or fluorides, oxygen, boron, and lithium.              2.5              Identify any conductivity excursions which may be indicative of resin                              intrusions.  Provide a technical assessment of each excursion and any                              followup actions.             2.6              Provide an assessment of the potential for any of these intrusions to                              result in a significant increase in the probability for IGA of VHPs                              and any associated plan for inspections.Required ResponseWithin 30 days of the date of this generic letter, each addressee is requiredto submit a written response indicating:  (1) whether or not the requestedinformation will be submitted and (2) whether or not the requested informationwill be submitted within the requested time period.  Addressees who choose notto submit the requested information, or are unable to satisfy the requestedcompletion date, must describe in their response any alternative course ofaction that is proposed to be taken, including the basis for the acceptabilityof the proposed alternative course of action.NRC staff will review the responses to this generic letter and if concerns areidentified, affected addressees will be notified.Address the required written reports to the U.S. Nuclear RegulatoryCommission, ATTN:  Document Control Desk, Washington, D.C. 20555, under oathor affirmation under the provisions of Section 182a, Atomic Energy Act of1954, as amended, and 10 CFR 50.54(f).  In addition, submit a copy to theappropriate regional administrator.The NRC recognizes the potential difficulties (number and types of sources,age of records, proprietary data, etc.) that licensees may encounter whileascertaining whether they have all of the data pertinent to the evaluation oftheir CRDM nozzles and other VHPs.  For this reason, the above time periodsare allowed for the responses.Related Generic Communications(1)                 Information Notice 90-10, "Primary Water Stress Corrosion Cracking                    (PWSCC) of Inconel 600," dated February 23, 1990.(2)                 NUREG/CR-6245, "Assessment of Pressurized Water Reactor Control Rod                    Drive Mechanism Nozzle Cracking," dated October 1994.(3)                 Information Notice 96-11, "Ingress of Demineralizer Resins Increases                    Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism                    Penetrations," dated February 14, 1996..                                                                                                                                                                                              GL 97-01                                                                                                                                                                                              April 1, 1997                                                                                                                                                                                 Page 9 of 10Backfit DiscussionUnder the provisions of Section 182a of the Atomic Energy Act of 1954, asamended, and 10 CFR 50.54(f), this generic letter transmits an informationrequest for the purpose of verifying compliance with applicable existingregulatory requirements.  Specifically, the requested information would enablethe NRC staff to determine whether or not the licensees' margins required bythe ASME Code, as specified in Section 50.55a of Title 10 of the Code ofFederal Regulations (10 CFR 50.55a) are met, that the guidance of GeneralDesign Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50,Appendix A, GDC 14) continues to be satisfied, and to ensure that the safetysignificance of VHP cracking remains low.  The requested information is alsoneeded to determine whether an augmented inspect- tion program, pursuant to10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety.Additionally, no backfit is either intended or approved in the context ofissuance of this generic letter.  Therefore, the staff has not performed abackfit analysis.Federal Register NotificationA notice of opportunity for public comment was published in theFederal Register  (61 FR 40253) on August 1, 1996, and extended on August 22,1996 (61 FR 43393).  Comments were received from seven licensees, two industryorganizations, and one Code Committee.  Copies of the staff evaluation ofthese comments have been made available in the public document room.Paperwork Reduction Act StatementThis generic letter contains information collections that are subject to thePaperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).  These informationcollections were approved by the Office of Management and Budget, approvalnumber 3150-0011, which expires July 31, 1997.The public reporting burden for this collection of information is estimated toaverage 80 hours per response, including the time for reviewing instructions,searching existing data sources, gathering and maintaining the data needed,and completing and reviewing the collection of information.  The U.S. NuclearRegulatory Commission is seeking public comment on the potential impact of thecollection of information contained in the generic letter and on the followingissues:  1.                Is the proposed collection of information necessary for the proper                    performance of the functions of the NRC, including whether the                    information will have practical utility?  2.                Is the estimate of burden accurate?  3.                Is there a way to enhance the quality, utility, and clarity of the                    information to be collected?.                                                                                                                                                                                                            GL 97-01                                                                                                                                                                                                            April 1, 1997                                                                                                                                                                                                            Page 10 of 10  4.                How can the burden of the collection of information be minimized,                    including the use of automated collection techniques?Send comments on any aspect of this collection of information, includingsuggestions for reducing this burden, to the Information and RecordsManagement Branch, T-6 F33,  U.S. Nuclear Regulatory Commission, Washington,DC 20555-0001, and to the Desk Officer, Office of Information and RegulatoryAffairs, NEOB-10202 (3150-0011), Office of Management and Budget, Washington,DC 20503.The NRC may not conduct or sponsor, and a person is not required to respondto, a collection of information unless it displays a currently valid OMBcontrol number.If you have any questions about this matter, please contact one of thetechnical contacts listed below or the appropriate Office of Nuclear ReactorRegulation (NRR) project manager.                                                                                                                                            signed by                                                                                                                        Thomas T. Martin, Director                                                                                                                        Division of Reactor Program Management                                                                                                                        Office of Nuclear Reactor RegulationTechnical contacts:  Keith R. Wichman                                                 (301) 415-2757                                                 E-mail:  krw@nrc.gov                                                 James Medoff                                                 (301) 415-2715                                                 E-mail:  jxm@nrc.govLead Project Manager:  C. E. Carpenter, Jr.                                                                 (301) 415-2169                                                                 E-mail:  cec@nrc.govAttachments:1.  Figure 1.  Typical Control Rod Drive Mechanism Nozzle