IR 05000336/1990008

Insp Rept 50-336/90-08 on 900425-27.No Violations Noted. Major Areas Inspected:Results of Eddy Current Exam of Steam Generators Performed During Oct 1989,w/emphasis on Acceptable Period of Operation for Steam Generators

ML20055G752
Person / Time
Site:	Millstone
Issue date:	07/02/1990
From:	Peter Habighorst, Lohmeier A, Murphy E, Strosnider J, Winters R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20055G750	List: IR 05000336/1990008 ML20055G749
References
50-336-90-08, 50-336-90-8, NUDOCS 9007240136
Download: ML20055G752 (13)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /90-08 L

Occket N License N DPR-65 Licensee: Northeast Nuclear Energy Company-

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P. O. Box 270 Hartford, Connecticut Facility Name: Millstone Nuclear Power Station - Unit 2 Inspection At: 3erlin, Connecticut Inspection Dates: April 25-27, 1990 Inspectors: (. M 9o l

h. W. Winters," Reactor Engineer, Materials & date P cess Section, EB,-DRS A. Lbhmeier, Reactor Engineer, Materials- &

6h o-date Processes Section, EB, DRS -

N P. Habighorst, Resident Inspector, Millstone, EL S o date DRP, Region I'

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4 44f90 E. Murphy, Se'TlioNEngineer, NRR, Materials date and Chemical Engineering Branch Approved by: c -

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7 s 4 J rosnider, Chief, Materials & Processes ' date St son, Engineering Branch, DRS, RI Inspection Summary: Announced inspection from April- 25-27, 1990--

(Report No. 50-336/90-08).

Areas Inspected: The result of the eddy current examination of the steam generators performed during October 1989 was inspected with emphasis on the dissenting opinion expressed by one member of th licensee's staff regarding -

the acceptable period of operation for the steam generator Results: No violations or deviations were identified. One unresolved item concerning the trending of primary-to-secondary leakage was identifie *

PDR ADOCK 05000336 Q PDC

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DETAILS 1.0 Persons Contacted Northeast Nuclear Energy Company

G. Alkire, Project Engineer, Piping Systems Engineering

F, Anderson, Senior Engineer, Nuclear Materials and Chemistry

P. Blasioli, Supervisor Nuclear Licensing

K. Colgan, Engineer, Nuclear Materials and Chemistry

D. Dube, Supervisor, Probabilistic Risk Analysis

- J. Ely, Supervisor, General Engineering

J. Fackelman, Project Engineer, Nuclear Materials and Chemistry

W. Hutchins, Millstone Unit 2, Licensing V. Jones, Project Engineer, Nuclear Materials and Chemistry

• • J. Keenan, Superintendent, Millstone Unit 2

M. Kupinski, Manager, Piping Systems Engineering

R. Linthicum, Engineer, Probabilistic Risk Analysis

R. Wells, Manager, Nuclear Materials and Chemistry United States Nuclear Regulatory Commission

W. Raymond, Senior Resident Inspector, Millstone

Denotes those attending the exit meeting in the Corporate Offic ** Denotes those attending the exit meeting at the site The inspectors also contacted other administrative and technical personnel during the inspectio +

2.0. Follow-up of Structural Integrity Action Plan Scope The scope of this portion of the inspection was to review licensee actions taken as a result of the October 1989 steam generator eddy current testing at Millstone Unit 2 with emphasis on the ' licensee's action plan for evaluation of the steam generator tube structural integrity. The results-of the October 1989 inspection and the status of the licensee's tube integrity evaluation were previously discussed at a meeting with NRC on February 22, 1990. At this meeting the licensee committed to perform additional testing and analysis of the cracked tubes removed from the steam

_ generators during the October 1989 inspection. .Also, as a result of newspaper articles reporting a differing opinion by an individual in the licensee organization, this inspection reviewed comments of this individual and the responses made by the licensee to determine if the individual's concerns were adequately addressed.

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2.1 Tube Evaluation I

~ i The licensee has performed an evaluation of the 86 crack profiles of tubes found during the October 1989 inspection using the ASME Code,Section III, '

Appendix F acceptance criteria. The results of this evalt.ation were that ;

85 tubes met these requirements and one tube (L45R13) did n,t after the '

170 effective full power days (EFPD) of operation prior to the October ;

1989 inspectio j j

l Independent review by a licensee consultant of these 86 tubes tsing different analytic techniques identified the same tube as not meeting Regulatory Guide 1.121 safety margins after the 170 EFPD of operation '

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i The calculated factor of safety for low probability, postulated accident f

conditions (faulted conditions) did not satisfy the R.G.1.121 criteri l l

Margin still existed against tube rupture under normal operating condition l

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Burst Testing Burst testing of tube samples was performed by the licensee to ascertain l the validity of the analytic methods used in predicting failure of tubes !

with varying crack geometries. Six tube samples containing manufactured ;

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flaws simulating actual crack geometries were pressure tested. Two tube '

l specimens removed from the steam generator were also pressure tested. Based !

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on these tests the licensee concluded that the analytical model was conservativ l 3-Dimensional Finite Element Analysis of Tube L45R13

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The analytical models used in the above tube evaluation did not explicitly !

consider crack ligament strengthening effects and tubesheet constraint. A !

i 3-D finite element analysis of the L45R13 was performed to assess the i effects. The results indicated that the.above analytical models were l conservative, but not over conservativ .2 Tube Life Calculations ,

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s l The licensee has concluded that.in the absence of an iniLiating event l (main steam line' break or main feedwater break) any tubes that fail will ;

exhibit leak-before-break behavior. Under these circumstances, the licensee ;

has concluded the existing leak detection systems will be adequate to allow l

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corrective action before a tube rupture. However, based on the eddy current examination performed in October 1989, calculations show that after 170 ,

EFPD there is a significant probability that stress in at least one tub could exceed the allowable level as required by Regulatory Guide 1,121.or "

ASME,Section III, Appendix F for faulted conditions. An independent analysis by a licensee contractor. indicated that the 170 EFPD was a conservative estimate of the time at which a tube would reach an

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unacceptable strt;s leve I s

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i A probabilistic assessment was made by the licensee's Probabilistic Risk ;

Assessment Organization on the basis of either 45 or 100 tubes susceptible to crack formation and propagation. The licensee estimated the probability !

of one or more tubes exceeding a stress of 83,800 psi (the stress causing rupture during an initiating event) for rupture ranges from 0.17 in j March 1990 to 0.91 in September 1990 for _ the assumption of 45 tubes and '

ranges from 0.34 to 1.0 over the same period for the assumption of 100 tubes. At the same time, however, the probability of an initiating event {

(e.g. steam line break) was estimated as being of a much lower order of '

magnitude. The combination of probability of tube rupture due to an '

initiating event was assessed in a licensee probabilistic analysis which indicated approximately a probability of 0.013 (between March and September 1990) for all cases of susceptiole tubes (45 or 100). Therefore, the licensee concluded that the probability of the tube rupture event combined with an event such as a steam line break is of such low magnitude as to justify continued operation through Septembe It should be recognized that this finding was tempered somewhat by the belief that tube rupture in the absence of an initiating event will be prevented by detection of a leak before brea ,

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2.3 Concerns Expressed by Individual With Dissenting Opinion i The following is a listing of concerns developed by a licensee engineer regarding the length of time for which the steam generators should be operated and the findings of the NRC inspector regarding these concern Concern 1

"Since the October 1989 in-cycle inspection was the-first time 100 percent "

UT sampling was performed of cracked tubes, the conclusion drawn in the first paragraph of page 6 (of the safety analysis) regarding the time dependency of depth of cracking is not technically supported. Currently applied techniques would indicate operation should be limited to less than 130 ef fective full power days (EFPD)." '

Inspectors Observations i

The inspectors discussed this concern with the individual. The individual stated that based on further testing and calculations subsequent to his )

original letter a revised estimate of the time for a projected tube to develop defects similar to the worst case found during the October 1989 inspection was 170 EFP Concern 2

" Traditional best estimate projections have underestimated the cracking '

problem as compared to the actual repair scope. The basis of 45 tubes should reflect historical cracking trends. This number should be around 100 to realistically determine a probability of a tube exceeding Regulatory Guide 1.121 requirements."

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l Inspectors Observations The inspectors discussed this concern with the individual. The individual is satisfied that Probabilistic Risk Analysis (PRA) performed by the licensee's PRA Department accounts for uncertainty in the number of tubes which may develop significant crackin !

Concern 3

"" Equivalent" depth should read "anrage" depth since time did not permit a !

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calculation of all 85 crack profi M to determine " equivalent" crack depth Based on a sampling of worst tubes the " equivalent" depth was on the-order I,

of 83 percent as compared to an average depth of 20 - 30 percent for the l- same crack and would result in a major change to the probabilistic model

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used in the safety assessment. The practice of using average depth as an indicator of, tube acceptability is good for uniformly degraded tubes (uniform depth, 360' circumferential extent); Sowever, it is an extremely poor method to assess tubes that challenge stability." i i

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l The inspectors discussed this concern with the individual. The individual !

l was satisfied with recalculations that were performed using.the " equivalent" depth that included a bending component in the calculations, These

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I calculations considered the basis of the actual geometry of the tube defect Concern 4

"The existing safety assessment assumes additional margins exist to rupture, based on data that is pressure based only. Potential external loads as a ;

result of tube bowing, tube lockup (from denting), secondary internals !

loading during accident conditions, etc., have not been-considered. An- '

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evaluation of these loads by Combustion Engineering for Millstone Unit N steam generators present conditions needs to be performed before we try j to strip our analysis of any potential conservatisms."  !

Inspectors Observations

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The inspectors discussed this concern with the individual. The individual

is satisfied that a 650 psi pressure load adjustment has been retained in j the analysis to account for these external loading effects. This value is i an estimate but calculations by Combustion Engineering determined that this value is conservativ Concern 5

"The limit load analysis performed for the most complex flaws is in a preliminary status with no independent review and approval being accomplished and forms a weak basis for subsequent safety assessment assumption Also, this analysis did not include Northeast utilities (NU's) standard practice of including NDE error in the analysis (a practice Regulatory Guide 1.121-also requests) or Section XI equivalent flaw length to depth criteria for multiple flaws."

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Inspectors Observations The inspectors discussed this concern with the individual and determined that the individual accepts that the limit load calculations have now been completed, reviewed and approved in accordance with the licensee's procedures. The interpretation of the analysis results has been upgraded to reflect additional test results, revised interpretation of flaw geometry, and improved analytic methods provided by the independent contracto l The individual has accepted the omission of inspection tolerance in the limit analysis because good correlation was found between the ultrasonic testing measurement and actual test value l l

Concern 6 '

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"NU has not performed a detailed Regulatory Guide 1.121 analysis for these tubes to date due to time constraints. Preliminary results identify that at least seven tubes will not meet this criteria."

Inspectors Observations The inspectors discussed this concern with the individual and determined that the individual agreed that Regulatory Guide 1.121 analysis had been performed in full by the time of this inspection. In the revised analysis in accordance with the Regulatory Guide 1.121 it was found that one tube did not meet the regulatory guidelines af ter 170 EFP Concern 7

"The present safety assessment identified a safety factor of 1.52, for tube L52/R22, between room temperature burst test information and the preliminary analytical result Regulatory Guide 1,121 requires the burst strength be determined experimentally at operating temperature. Margins may not be presen Evaluation of end of cycle (E0C) 8 pulled tube L25/R19, which leaked at 0.1 gpm in service shows zero margin between the analytical model and actual rupture poin .

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Inspectors Observations The inspectors discussed this concern with the individual and determined that the individual believes it is acceptable to use room temperature test for determining rupture properties at operating temperatures because there are only small differences between the properties at these temperatures, i

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More recent testing has shown safety factors greater than 1.24 between burst test results and analytical predictions, compared to the 1.52 rnargin from initial testin i Regarding tube L25/R19, the crack may have exhibited further growth after it was plugged, but prior to its being removed from the field. The '

individual does not consider the size of the crack to have direct implications ter -the current operating cycle,

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Concern 8

"The minimum calculated wall thickness of 77% was based on a flow stress of 69,000 psi. This flow stress was assumed to be conservative since this was considered to be a code based (3 Sm) number. A detailed review of Millstone Unit No. 2 certified material test reports identified that twenty-nine heats have flow stresses below this valu The minimum flow stress is 65,000 psi and results in a minimum wall thickness of 75%. This minimum wall. thickness was not being utilized in the safety assessment submitted to the NRC prior to the February 22, 1990 mecting. The number determines a statistical interval used to arrive at the current ufety '

assessment's conclusion."

Inspectors Observations The inspectors discussed this concern with the individual and determined that the individual based his question regarding the results of analysis and testing on information-available at the time of his question. Since that time however, new considerations and additional tests, analyses, and evaluations, have resulted in a somewhat different safety assessment with which he agree '

Concern 9

"A review of tubes with uniform through wall cracks over 360 of the i circumference was performed with a comparison to the 77 percent acceptance criteri This review was not performed for the new 75 percent acceptance criteria due to time constraints."

Inspectors Observations The inspectors discussed this concern with.the individual and determined that the individual agrees with the method presently in use that includes provision for including total stress (with bending).

Concern 10

"E0C 8 leaking tube L25R19 exhibited a circumferentially predominate degraded region of 190 which resulted in a through wall crack of 35 and a leak of 0.1 gpm under normal operating conditions. The tubes presently being evaluated _have regions greater than 190 and are VT determined to be 80%

through wall. Longer term operation of tubes similar to these could provide for through wall cracks greater than 0.1 gpm at normal operating condition rr

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Inspectors Observations The inspectors discussed this concern with the individual and determined that the individual agrees that as a result of the testing and evaluations the plant can operate for 170 EFPD before reducing the safety margin below Regulatory Guide and ASME limit I 2.4 Conclusions Based on discussions with the individual who had expressed concerns about the length of time the steam generators should be operated, the inspectors concluded he now believes his concerns have been fully addressed and given due consideration by his management. Sased on discussions with the engineer having the dissenting opinion and after reviewing the results of the investigations made by the licensee as a result of this dissenting opinion the inspectors concluded that the 10 concerns had been adequately addrcsse The licensee has p;rformed a probabilistic assessment of tube rupture under faulted conditions in justification of operation through the end of the fuel-cycl Included in their assessment is the assumption that-a leaking tube can be identified during normal operation. On the basis of the calculations-and testing performed on tubes removed from the steam generators, the licensee stated that the conclusions expressed in the February 22, 1990 meeting are still valid and operation to the end of cycle 10 is justifie .0 Primary-to-Secondary Leakage Monitoring Scope A review of available licensee radiation monitoring and radiochemistry sampling process to detect primary-to-secondary leakage was performed to assess the adequacy of the licensee's program. Specifically, the inspector

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reviewed detection methods, procedures, administrative controls, and operator actions in response to applicable control-room alarms, radio-chemistry results, and licensee management awareness to the_ leakage progra .1 Detection Methods The radiation monitoring instruments used at Millstone 2 to detect leakage from the primary-to-secondary are: (1) steam jet air ejector radiation-monitor (RM-5099); (2) steam generator blowdown radiation monitor (RM-4262)

(3) N-16 radiation monitors; (4) main steam line radiation monitors (RM-4299 A,B,C) and (5) alternate steam jet air ejector monitor. The radiochemistry analysis includes steam generator blowdown grab samples, secondary tritium calculations, 'and steam jet air ejector grab sample <

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The steam jet cir ejector radiation monitor (RM-5099) is located at the 31' 6" elevation of the turbine building. The monitor's required operability l 1s based on technical specification (TS) 3.3.3.9. The design basis is to monitor noble gas activity in the condenser air removal system and to provide a control logic signal to close the steam generator blowdown valves 1 to terminate discharge of blowdown to the circulating water canal. The control logic signal is to prevent 10 CFR 20 radioactivity limits from being exceeded to unrestricted areas. The licensee considers RM-5099 to I be one of the first monitors to sense primary-to-secondary leakag The steam generator blowdown radiation monitor (RM-4262) is located on the 14'6" elevation of the auxiliary buildin !

RM-4262 monitors a combined '

blowdown activity from both steam generators. The monitors operability is based in TS 3.3.3.9. The design basis is similar to RM-5099, except for the monitored location point. The licensee's radiological assessment branch '

review of RM-4262 indicates it is not a sensitive indication of primary-to-secondary leakage based on industry experience. Specifically, in actual past industry tube rupture events it often took up to 30 minutes for activities in the steam generator water to buildup to cause a monitor alar The N-16 monitors a're located on each main steam.line on the 38'6" elevation of the turbine building. The monitors were installed in July, 1989 under licensee inservice test T89-019. The licensee determined within the inservice test no unreviewed safety question per 10 CFR 50.59 existed. This monitor was installed as an independent action by the licensee and is not a-regulatory requirement. Therefore, it is not-required to be operable'nor is it identified in the Final Safety Analysis Repor The monitors-detect the presence of gamma radiation in the main steam lines. The digital leak rate calculations are based on high energy gamma radiation from the radio-active decay of Nitrogen-16 and the total counts per second readout display the detected low energy gamma. The monitors detect the presence of a steam generator leak, and identify the affected steam generator unlike RM-5099 and RM-4262 that do not isolate the source of the. leakag The main steam line radiation monitors (RM-Gs A,B,C). are located in the 38'6" east and west penetration rooms. The monitors are located on the main steam lines to detect primary-to-secondary leakage and quantify a postulated post accident release thru the main steam line safety valves and atmospheric dump valves. The operability of the monitors is prescribed in TS 3.3.3.8 and NRC Regulatory Guide 1.9 The alternate steam jet air ejector monitor was installt on April 12, 1990 under authorized work order M2-90-0369 The monitor is iocated upstream

- 1 of RM-5099, at the 14'6" elevation of the turbine building. The monitor's s primary function is a backup and verification of monitor RM-5099. No design basis or requirements exist for the alternate steam jet air ejector monito I .

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i The radiochemistry input into detection of activity in the steam generators is in part required by the technical specifications and further amplified by procedures and lictnsee administrative guidance. TS surveillance requirement 4.7.1.4 specifies the time interval for gross activity and .

dose equivalent iodine in the steam generators, and TS 4.11.1.1. radioactive liquid samples levels not to exceed 10 CFR 20 Appendix B, Table II, Column 2 for release from the sit F Procedurally, per SP-2833. " Secondary Coolant Ar.dlysis for Total Activity",

the licensee samples steam generator blowdown grab samples and air ejector grab samples daily, and upon c RM-5099 or RM-4262 alarmed conditio In addition, the frequency is increased based on a calculated primary-to-secondary leakage. (i.e. SJAE grab samples three times per day when leakage exceeds .05 gallons per minute).

The leakage rate calculations are based on three methods identified in licensee procedure CP-2806Y. The calculation methods are based on Electric >

Power Research Institute (EPRI) NP-5960-SR " Primary Water Chemistry ;

Guidelines." The methods implemented by the licensee consist of a tritium balance, air ejector gas method, and soluble liquid nuclei determination of steam generator leakage. The licensee employs the tritium analysis as a confirmation and backup to-the air ejector gas method for leakage rate determinations, The air ejector gas method is the primary method implemented

.by the licensee. This method provides a rapid response to leak rate change'., ,

however, the detection sensitivity is significantly less than the triti samples, based on detectability of liquid scintillation vice the gamm spectrometr The disadvantage of the tritium method is a relatively slow response in analysis acquisition and determination of time constan The licensee calculates, as a minimum, a daily leak rate calculation using the air ejector gas metho .

3.2 Procedures and Administrative Controls l

The. inspector reviewed licensee's procedures and actions for a primary-to-secondary leak to assess the adequacy of required actions at the facilit Procedures reviewed are listed below:

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OP 2260 " Millstone 2 Emergency Operating Procedure (EOP) Users Guideline."

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AOP 2569 "SG Tube Leak"

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E0P 2534 "SG Tube Rupture"

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SP 2833 " Secondary Analysis for Total Gamma-Activity"

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CP 2806Y " Calculating Primary to Secondary Leak Rates" )

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OP 2316A Section 6.15 " Main Steam System" ,

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Plant Operations Night Order 7/14/89-1

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OP 2383A " Process Radiation Monitors Operation" l Inspector review identified adequate required operator response to alarms for the diagnostics of a primary-to-secondary leak, increased radiochemistry sampling in response to. increases in leakage, and the licensee's process to evaluate radiochemistry data. In AOP 2569, the decision to continue plant operation with a known leakage resides with plant management, and if technical specifications for leak or SG activity are exceeded, a required shutdown of the plant is reinforced oy procedure and in accordance with the applicable action statemen The present licensee procedures do not provide trending of leakage rates to take plant operational ection prior to exceeding required TS limits or onset of a postulated steam generator tube rupture. The above enhancement program was addressed for information to Millstone 2 via NRC Bulletin 88-02 dated February 5, 1988. The inspector considers this issue open (90-08-01)

and will follow licensee actions to this regard in future inspection .3 Operator Actions to Applicable Radiation Monitor Alarms and Radiochemistry Results

! Inspector observations of control room activities noted that, once RM-5099 l or RM-4299 alarms and isolates the steam generator blowdown, control room

! operators request a chemistry department sample of either the steam jet air ejector or blowdown. The control room logs document-the time of the alarm, automatic action, chemistry radioactivity samples, and N-16 radiation monitor readings. The above actions are in accordance with procedure l OP 2316A and plant operations night order 7/14/89- Control room alarms available to provide indication of a primary-to-secondary leak are: " Main Steam Line High Radiation / Instrument Failure," Process Monitor Radiation High and local <.larms at the RM-4262, RM-5099, and the N-16 monitor .4 Licensee Management Awareness

. Inspector observation at licansee daily work planning meetings indicates heightened awareness of leak rate calculations on a daily basis, and '

control room operators are cognizant of current leakage values. Utility management actions pursuant to current leakage rates include (1) implementa-tion of alternate air ejector monitor (2) confirmation of leakage rates by

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tritium analysis and predictions (3) ratics of noble gas half-lifes to determine an approximate tin.e delay between primary and secondary systems, and (4) evaluation of air ejector radiation " spikes".

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3.5 Conclusions No deficiencies were observed in the licensee's primary-to-secondary leak rate monitoring program. The program includes monitoring capabilities not required in TS or other regulatory requirements and additional sampling actions by the chemistry department (i.'e. tritium predictive analysis).

4.0 Water Chemistry Millstone Unit 2 receives makeup water from the New London, Connecticut city water supply. This water is suitable for human consumption but contains impurities detrimental to the operation of the steam generator The licensee treats the city water by filtering and ion exchange to remove most of the impurities. The next step is to use reverse osmosis to remove organic materials peculiar to the area, then another ion exchange for final treatment. The treated water is then stored in a stainless steel tank for later use. Makeup water is fed into the secondary system through the 'A'

hot well of the main turbine condenser as needed to replace blowdown losse This serves to remove any dissolved _ oxygen or other noncondensible gases through the steam jet air ejector From the condenser hot wells the flow is through the condensate pumps, polishers, feedwater heaters, feedwater pumps, to the steam generators, as steam to the turbine and back to the condensers. In line chemical monitoring is done before and after the condensate pumps, and on the steam generator blowdow Radiation detectors to monitor for steam generator tube leaks are at the steam jet air ejectors, and on the steam generator blowdown lines. In addition there is an N g radiation monitor located in the main steam line between the steam generator and the_ turbine. Water flow through this system is'on the order of 18,000 gallons per minut Chemical additions of ammonia and hydrazine are made downstream of_the condensate polishers to adjust the pH and control oxygen respectivel These additions are made by continuous, automatic injection of solutions to the main flow. A boric acid solution is also injected _ downstream of the-polishers by manual control of the injection pump. The boric acid is added to the secondary water in an attempt to eliminate the cracking the steam generator tubes have experienced at the top of the tubeshee During transients, the water chemistry undergoes changes. During these transients, special attention is required to maintain the chemistry at appropriate levels, Conclusions "

The licensee has developed an excellent system for the control of water chemistr Purity is approaching the-current technological limit of the ability to measure the impurities presen u

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5.0 Unresolved Items j

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Unresolved items are matters about which more information is required in

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order to ascertain whether they are acceptable items or violations, j Unresolved items are discussed in. paragraph '

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6.0 Management Meetings  :

l Licensee management was' informed of the scope and purpose of the inspection i at the entrance interview on April 25, 1990. The findings of the inspection !

were discussed with licensee representativas during the course of the !

inspection and presented to licensee management at the April 27, 1990 exit {

interviews (see paragraph I for attendees), h At no time during the inspection was written material provided to the ,

licensee by the inspector. The licensee did not indicate that proprietary !

information was involved within the scope of this inspection, j I

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