ML20217C701

Responds to Expressing Concern Re Ref Temps for Plant Reactor Vessel Exceeding Screening Criteria for Pts. NRC Has Evaluated Vessel Issues & Concluded That Vessel Conditions Continues to Provide Adequate Protection

ML20217C701
Person / Time
Site:	Yankee Rowe
Issue date:	07/10/1991
From:	Murley T Office of Nuclear Reactor Regulation
To:	Hodgkins C MASSACHUSETTS, COMMONWEALTH OF
Shared Package
ML20217C706	List: ML20217C701 ML20217C716
References
2.206, NUDOCS 9107180085
Download: ML20217C701 (3)

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July 10, 1991 Docket No. 50-029-The-Honorable Christopher J. Hodgkins The Commonwealth of Massachusetts State House Boston, Massachusetts 02133

Dear Mr. Hodgkins:

'I am' responding to your letter of June 19, 1991, in which you expressed concern over the Yankee Rowe Nuclear Power Station in northwestern Massachusetts.

On August 31, 1990, the Nuclear Regulatory Commission (NRC) staff issued a safety assessment of the Yankee Rowe reactor vessel (copy enclosed) and concluded that there was reasonable assurance that-the facility-could be

. operated for an additional operating cycle, currently expected to be completed in early 1992.

In your letter, you indicated a concern that reference temperatures for the Yankee Rowe reactor vessel exceed the screening criteria -for pressurized thermal shock (PTS)~ screening as delineated in Section 50.61(b)(2) of Title 10 of the Code of Federal Regulations (10 CFR 50.61(b)(2)). The NRC may, as' specified in 10 CFR 50.61(b)(5), on a case-by-case basis, approve operation of a. facility _ at values of reference temperatures in excess of the PTS screening criteria. The rule requires the ctaff to consider factors significantly affecting: the-possibility for failure _of; the reactor vessel-including the results of a probabilistic fracture mechanics analysis in determining to approve operation.

In-its August 31, 1990 safety assessment,.the NRC staff assessed the_results;from; Yankee Rowe's probabilistic fracture mechanics analysis and'

.found them_ acceptable.

As you may be aware, on June 4, 1991, the Union of Concerned Scientists (UCS)~and

.the New England' Coalition on Nuclear Pollution (NECNP)-petitioned the NRC, pursuant to the provisions of Section 2.206 of Title 10 of the Code of Federal

~

. Regulations._(10 CFR-2.206), toliamediately shut down Yankee Rowe.

By.-letter-of June'25, 1991-(copy enclosed), the Director of-the Office of Nuclear Reactor

-; Regulation responded to the Petitioners, stating.that the__NRC. staff had found that the'_ Yankee Rowe-reactor vessel-does not pose.an undue risk to the public healthand safety. Accordingly, the Director, determined =that Petitioners' concerns did not warrantiimmediate action to shut down Yankee Rowe.

Consistent with 10 CFR 2.206, the NRC will further_ address the specific issues raised by

-the_ Petitioners,.and the staff is. preparing the detailed response.

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'The'Nonorable Christopher J. Hodgkins July 10, 1991 The Commission has-scheduled a meeting on July 11, 1991, at its Rockville, Maryland office to discuss the reactor vessel issuest In-addition, the. staff will conduct a public_ meeting with the licensee in the Rowe, Massachusetts, area in the near future. Following the meeting, the public will be afforded the opportunity to address their questions and concerns to the NRC staff.

I assure you the NRC staff has evaluated the Yankee Rowe vessel issues carefully-and has concluded that the vessel condition continues to provide adequate protection of the public health and safety.

Sincerely, original signed by J. Partlow for Thomas E.- Murley, Director

' Office of Nuclear Reactor Regulation

Enclosures:

1.

Safety Assessment dated 8/31/90 2.

Ltr'to D. Curran fm T. Murley dated 6/25/91 cc w/ enclosures:

Mr._ George Papanic,_Jr.

Senior Project Engineer - Licensing

-Yankee Atomic Electric Company 580 Main Street Bolton,-Massachusetts 01740-1398 See'next page

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,AUG 3 g 1990 Docket No. 50 029 Dr. Andrew C. Kadak President and Chief Operating Officer Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398

SUBJECT:

YANKEE R0WE REACTOR VESSEL By letter of July 5,1990, you submitted for staff review the report, " Reactor Pressure Vessel Evaluation Report for Yankee Nuclear Power Station." This report was in response to our letters of May 1, 7, and 15,1990. Because of our concerns regarding reactor vessel integrity, we requested information that was needed to assess the effect of vessel operating temperatures, beltline material chemical composition and material surveillance test results. These concerns can affect the conclusions of previous NRC reviews of vessel integrity.

Those previous reviews considered postulated Low Temperature Over Pressurization s

(LTOP) events Pressurized Thermal Shock (PTS) events, and low irradiated Charpy Upper-Shelf Energy (USE).

In your July 5,1990, submittal, you stated that the RTNDT values for reactor vessel plate and. weld metal for the years 1990 and 2000, are below the screening criteria of 270'F and only slightly above the screening criteria for the year 2020'. Additional-information was provided to support your statements in numerous comunications which are listed as references in the attached NRC safety assessment report.

The staff, in its review of your submittal, has concluded that there are sub-stantial uncertainties associated with the weld chemistry and the effects of coarse grain plate material on the shift in-the RTNDT reference temperature.

These uncertainties could result in reference temperatures significantly higher than the screening criteria specified in the regulations.

However, staff calculations,. recognizing these uncertainties coupled with este ates of the likelihood of the occurrence of PTS events, lead us to conclr e that it is

-acceptable to operate the Yankee Rowe plant until the end o' fuel cycle 21 (approximately February 1992).

Although.your July 5,1990 submittal did not consider LTOP events, the staff has evaluated this scenario based upon additional information provided by your staff. We consider that the systems and procedures implemented at Yankee Rowe, along with estimates of vessel conditional probability of failure provide

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Dr. Andrew C. Kadak AUG 311930 sufficient assurance that the probability of an LTOP event leading to brittle vessel failure is sufficiently low to permit continued operation for an additional cy c l e.

In your July 5,1990, submittel, you stated that af ter performing an additional analysis using the ASME Section XI methodology the USE calculated for the Yankee irradiated plate and weld are 35 ft-lb and 40 ft-lb respectively.

Paragraph IV.A.1 of Appendix G, 10 CFR 50 states that reactor vessel beltline materials must have USE throughout the life of the vessel of no less than 50 ft-lb, unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation, that lower values of upper shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of the ASME Code. According to staff calculations the USE for the Yankee Rowe vessel could be as low as 35.5 ft-lb. Your USE analysis indicates that the reactor vessel with 35 ft-lb Charpy USE has margins of safety against fracture, equivalent to those in Appendix G of the ASME Code as required by the regulations. The staff has reviewed your analysis and considers that it is acceptab'.e to operate the Yankee Rowe reactor vessel until the end of fuel cycle G with 35 f t-lb Charpy USE.

During a meeting on August 21, 1990 you agreed to provide us within 60 days 'of the meeting date, a proposed plan to addres= the uncertainties noted herpin.

The enclosed staff safety assessment addresses these uncertainties in detail.

For Yankee Rowe to continue to operate beyond the next operating cycle, we stress the need for you to reduce the uncertainties in the various elements affecting reactor vessel integrity.

Your plan should include any appropriate procedural changes, technical specification changes, and sampling and physical tests to ascertain the chemical and physical properties of reactor vessel lower plates and welds.

The staff concludes the following actions should be included in your plan:

Long Term Actions to be Completed prior to Cycle 22.Startup 1.

Develop inspection methods for the beltline welds and each beltline plate from the clad to 1 inch from the clad / steel interface to determine if the metal contains flaws.

Perform tests on typical Yankee Rowe base metal (0.lg-)0.207, Cu) to 2.

determine the effect of irradiation (f = 1-5E19 n/cm

, austentizing temperature (1650*F-1800*F) aad nickel composition (0.18-0.70 percent) on embrittlement at 500*F and 550'F irradiation temperatures.

3.

Determine composition of the circumferential weld metal in beltline by removing samples from the weld.

In addition, prior to Cycle 22 startup, you should install surveillance capsules in accelerated irradiation positions. The capsules are to inclJde materials representing the beltline circumferential weld metal and upper and lower plates.

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Dr. Andrew C. Kadak AUG 3119%

In addition, you also agreed to the following:

1.

Fluence calculations prepared by Westinghouse will be provided by October 1, 1990.

2.

Results of peer evaluation of Yankee's July 5, 1990, submittal will be proviced within three months.

The results of the staff's review of your July 5,1990, submittal are included in the enclosed safety assessment.

Sincerely,

" v.-? - Y Thcras E. Murley, Director Office of Nuclear Reactor Regulation

Enclosure:

As stated cc:

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Dr.; Andrew C. Kadak AUG 311990

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'i in addition, you also agreed to-the following:

1.

Fluence calculations prepared by Westinghouse will be provided by October 1, 1990.

2.-

Results of peer evaluation of Yankee's July 5, 1990, submittal will be provided_within three months.-

i The results of the staff's review of your July 5, 1990, submittal are incliided in the enclosed safety assessment.

i Sincerely, Original !gned byl necas I. Kurley Thomas E. Merley, Director Office of huclear Reactor Regulation

Enclosure:

As stated cc:

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0FFICIAL RECORD COPY d

,e Document Name:

YANKEE R0WE REACTOR VESSEL

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O Or.-Andrew C. Kadak cc:

Thomas Dignan, Esquire

-Ropes and Gray 225-Franklin Street-Boston, Massachusetts 02110 Mr. T. K. Henderson Acting: Plant Superintendent Yankee' Atomic Electric Company Star Route Rowe, Massachusetts 01367.

Resident Inspector Yankee Nuclear Power Station U.S. Nuclear Regulatory Commission Post Office Box 28 Monroe Bridge, MassachusettsL 01350

-Regional Administrator, Pegion I U.S. Nuclear Regulatory Commission 475-Allendale Road King of Prussia, Pennsylvania 19406 Robert M.-Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 7th Floor.

Boston, Massachusetts 02111 Mr. George Sterzinger Commissioner Vermont Department of Public Service-120 State Street, 3rd Floor l

Montpelier, Vermont 05602' Ms. Jane M. Grant Senior Engineer - License Renewal Yankee Atomic Electric Company l-580 Main Street l

Bolton, Massachusetts 01740-1398

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ENG05URE SAFETY A$5ES$ PENT OF YANKEE R0WE VESSEL I.

INTRODUCTis In a letter dated July 5,1990 from John D. Haseltine, the Yankee Atomic Electric Company (the licensee) submitted for staff review a report entiticd,

' Reactor Pressure Vessel Evaluation Report for Yankee Nuclear Power Station."

The report was in response to NRC letters dated May 1, 7, and 15, 1990.

The

- staff letters requested additional information, which was needed to assess the effect of vessel operating temperatures, beltline material chemical composition, and material surveillance test results on the integrity of the Yankee Rowe

- reactor vessel.

These concerns have potential impact on prior NRC reviews of vessel integrity resulting from low irradiated Charpy Upper Shelf Energy (USE) and vessel' integrity during postulated Pressurized Thermal Shock (PTS) and Low TemperatureOverpressurizationEvents(LTOP) events.

The licensee's justification for operation of Yankee Rowe is that there is adequate assurance that risk of vessel brittle failure is very low.

This conclusion depends upon two factors:

(1) the-frequency of challenges to the

. vessel, and (2) the probability of vessel failure given a challenge. event (conditionalvesselfailureprobability).

Brittle failure challenge events L

fall into 2 general categories:

(1) pressurized thermal shock (PTS) events, and (2)' low temperature overpressurization (LTOP) events.

For both categories the licensee has estimated a very low probability that a vessel failure will

occur.

The frequency of challenge-and probability of vessel failure for PTS I

and LTOP events are-discussed in Section 11 and Section III respectively.

Additional information to support the licensee's conclusion was submitted in References 12 through 23.

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2 II.

PRESSURIZED THERMAL SHOCK (PTS) EVALUATION t

II.1 Systems Evaluation of PTS Limiting Events for PTS events the licensee has indicated that risk contributors can be divided into 3 groups:

(1) steam line breaks, (2) small break LOCAs, and (3) transients.

For each PTS group the event resulting in the most limiting temperature and pressure conditions (from a vessel failure perspective) is considered to be representative for the group.

The frequency for a group is the sum of the frequencies for each event in the group.

For PTS the staff's review focused on the following conside-ations:

(1) completeness of the events considered; (2) the adequacy of the thermal hydraulic analyses; (3) adequacy of the event frequency es ' mates including human error contributions; and (4) adequacy of the lir,ttint events selected.

II.1.1 Comoleteness of PTS Events Considered In its PRA submittal on PTS for Yankee Rowe, the licensee performed a systematic evaluation of initiating events (IEs) that could lead to primary system overcooling coupled with primary system repressurization. These IEs were grouped into four major categories.

Category I is main coolant system (MCS) induced events.

This category of events includes HCS - initiated cool-down events, depressurization events, and injection events, with both the MCS intact and faulted.

Category II is secondary system induced events. This category includes events initiated due to steam removal, feedwater flow, steam generator blowdown, and steam /feedwater flow control abnormalities.

Category III is general transients which do not directly result in initial MCS cooldown and are not related to support systems but, if followed by other system failures could result in cooldown events.

Category IV is events not necessarily resulting in initial MCS cooldown but involving support' systems which have the potential to impact other frontline systems which could cause MCS cooldown. The licensee also reviewed the PTS evaluations for H. B.

Robinson and Calvert Cliffs performed by Oak Ridge National Laboratories to e

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assure that the Yankee Rowe evaluatic.. took into account sequences found to be significant contributors to thermal shock at these plants. The licensee examined the operating experience at Yankee Rowe (including all the trip logs) and concluded that there has never been an overcooling event at Rowe.

The plant design and the Yankee Rowe Probabilistic Safety Study were likewise reviewed to identify any plant unique cooldown sequences.

11.1.2 Thermal-Hydraulic Analyses for Transients Affecting PTS Based on system and thermal-hydraulic considerations, each of the initiating events were evaluated and the initiators relevant to PTS concerns were identified.

Event tree sequences were then developed for each event associated with the relevant initiators concerning PTS.

Support systems were treated in a separate auxiliary tree. Quantification of event sequences and endstates was performed based on the system models, dependencies, and human actions..

Endstates with frequencies higher than 10-8/ reactor year were selected for potential further thermal-hydraulic and fracture mechanics analysis.

Bas'ed on grouping sequences with similar plant thermal-hydraulic behavior, this process resulted in the final set of initiating events being grouped into three categories with four corresponding event trees:

steam line breaks upstream or downstream of non-return valves, small break 1.0CAs, and transients.

For each of the above identified four event trees, thermal-hydraulic analyses were performed to model the spectrum of overcooling events.

The transient downcomer temperature and MCS pressure were calculated and bounding cases affecting PTS concerns were identified.

The licensee used the CEPAC computer code to perform scaping calculations for the events of concern to predict limiting cooldown transients at Yankee Rowe.

Based upon pressure and temperature response a small break LOCA of 15/16 inches at the reactor coolant pump suction and three cases of main steamline break were found as the limiting transients relative to PTS concerns.

These l

limiting transients were analyzed in greater detail using the RETRAN computer l

code, the combination of RETRAN and EPRI models, or the combination of RETRAN l

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4 and REMIX-codes. - The RETRAN computer code is designed to analyze the response of plant systems during both normal and transient conditions. The licensee's capability of_using RETRAN for main steam line break analyses was reviewed and cpproved by the staff in 1983.

The licensee asserted that the CEPAC code is similar but simpler than the RETRAN code.

The CEPAC code has not been reviewed by the staff.

However, the limiting transients results were not based upon

-CEPAC calculations. The-EPRI model has been used for the non-stagnant ficw conditions in the Calvert Cliffs PTS analysis.

The REMIX code was used for the SBLOCA case without offsite power available, where flow stagnation occurred.

-The-staff has evaluated the adequacy of the licensee's use of REMIX for the Yankee Rowe plant SBLOCA case.

We feel that sufficient conservatism exists in this analysis

~There are other conservative assumptions considered in the SBLOCA analysis such as early stagnation in the downcomer area, low decay heat, coincident loss of-offsite power and an assumption that all three trains of safety injection are injecting water to the MCS. The first three of these assumptions result in minimal mixing of the cold SI water with the hot primary system water.

The fcurth assumption maximizes the amount of cold water added to the primary system. Theresultisaconservative(colder)downcomerwater temperature.

In the main steam line break cases, there are conservative assumptions applied such as zero power at event initiation, low decay heat, dry steam to the break, coincident loss of offsite power, non-return valve failure.

etc.

As in the SBLOCA cases these assumptions minimize mixing in the primary system and maximize primary system cooldown.

The following design features were-found to be significant in the analyses:

The charging pumps trip on a safety injection signal.

This feature helps assure that-the maximum repressurization achievable during a LOCA or transient that may initiate safety injection is limited to the shutoff

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head of the safety -injection system (1550 psig).

The safety injection pumps have relatively low capacity and a shutoff head of 1550 psig when HPSI and LPSI are aligned in series. When not aligned in series the shutoff head is limited to 800 psig.

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i There is only one pressurizer PORY. This reduces the probability _ of a stuck open PORV.(relative to two PORVs) initiating a cooldown event.

There is onlyLone turbine bypass valve, and it has low capacity. _ This tiimits'the rate of potential cooldown (if the valve fails open).

The emergency atmospheric steam dump valves have low capacity. This t

limits the rate of.cooldown should the valves fail open.

The condensate pumps trip following a steam break in the vapor containment.

-: 'i Emergency feedwater pumps must be manually started.

l

.The Emergency Operating Procedures direct the operator, in response to iminent PTS conditions, to stop safety injection pumps and low pressure safety injection pumps if there is sufficient subcooling and pressurizer.

level.

Although the plant has primary system loop isolation valves, emergency L

_ operating procedures only require their operation during a steam generator tube rupture in order to isolate the faulted generator.

For other LOCAs I

inside the vapor' containment, the operators are-instructed to not isolate the break location.

Isolation of a break could ~ result in significant-repressurization.

The feedwater pumps trip on reactor scram or low suction pressure. Above 15 percent power, operators are instructed to isolate feedwater flew by

closure of the feedwater regulating valves and the'feedwater motor-operated i: solation' valves. These measures limit the chance _and severity of an overcooling event caused by overfeed of the steam generators.

Yankee Rowe:is also_ unique in the large number of ways in which water can be supplied to the steam generators. Among-these multiple paths, all flow sources however, are dws.rfed in volume by the boiler feedwater pumps. The feedwater

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control system hac independent controls for each steam generator such that a single failure in the control system would not result in overfeeding more than i

one steam generator.

If another system should begin to supply additional water to the steam generators (e.g., the charging system), the feedwater control system would cut back on the flow from the boiler feedwater pumps to maintain steam generator level, in view of these plant specific features and the modelling assumptions used by the licensee, the staff considers that the thermal-hydraulic analyses are conservative and reasonable. We note that the results are also consistent with other similar anelyses such as the Robinson and Calvert Cliffs PTS studies.

!!.1.3 Frequency of Cooldown Events Threatening the Vessel Yankee Atomic has estimated that the frequency of sequences that would significantly challenge the integrity of the reactor vessel due to pressurized thermal shock to be about 5-E-4 per reactor year.

Small break LOCAs result in the most limiting thermal hydraulic conditions of any of the sequences analyzed.

Yankee Atomic estimated this frequency by partitioning the WASH-1400 small break LOCA frequency (for break sizes between 0.5 and 2 inches) based on the number of pipe segments inside the vapor containment that were between 1 l

and 2 inches in interior diameter (l.D.).

The limiting sequence (combination of frequency and thermal hydraulic conditions) was estimated by Yankee to be a l

LOCA about 15/16 inches I.D. where the estimated minimum downcomer temperature was 151'F and the maximum RCS pressure after cooldown was 670 psi.

This analysis did not, however, taka into account the possibility of the operator violating his Emergency Operating Procedures and attempting to isolate the break.

Such action could lead to an RCS maximum pressure equal to the shutoff head of the safety injection pumps.

At the staff's request Yankee Atomic performed an analysis of such a sequence.

The licensee concluded that it was not a significant event because of the small amount of small bore piping which is isolable, the frequency of a small break in any location, and the operator training and procedures which direct operators not to isolate breaks inside the vapor containment.

7 The staff has reviewed the licensee's event frequency estimates in considera-tion of the piant specific features of Yankee Rowe.

The limiting event frequencies are reasonably consistent with values used in other studies.

The treatment of human error in the Yankee Rowe PTS PRA is judged to be conserva-tive or non-conservative depending on the timing of the error. The PTS thermal hydraulic analyses indicate that small break LOCAs give the worst combination

- of low primary system temperature, high primary system pressure, and high The staff believes that the licensee's estimate of 5x10"# per cooldown rate.

reactor year as the frequency of a small break LOCA is consistent with the frequency of 1x10 per reactor year typically used in PRAs.

11.1.4 Adecuacy of PTS Limitino Events The licensee performed a systematic review of the Yankee Rowe features in order to identify potential overcooling sequences.

The licensee then grouped the possible events on the basis of similarity in thermal hydraulic (TH) response; Fcr each group a limiting event was determined based upon consideration of event frequency and the severity of pressure temperature conditions (relative to vessel failure) resulting from the event.

The staff concluded the events considered are reasonably comprehensive, the thermal hydraulic analyses, methods, assumptions and results are reasonable. With regard to the frequency estimates, the most important considerations are the insensitivity to human error and the relative frequency values.

The systems failure estimates used are considered to be reasonable because they are consistent with state-of-the-art PRA applications. The event frequencies were also found to be relatively insensitive to human error since the limiting events would not change significantly even if the human error probability (at times greater than l

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) changed by a factor of 100. Therefore, based on these systems, thermal hydraulic, and event frequency studies, the staff concludes that there is

- reasonable assurance that the limiting events have been properly identified.

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11.2 PTS Materials Evaluation

-i II,2.1 Background

• The Pressurized Thermal Shock (PTS) rule,10 CFR 50.61, adopted on July 23, 1985, establishes a screening criterion that is a limiting level of embrittle-rnent beyond which operation cannot continue without further plant-specific evaluation. The screening criterion is given in terms of RTNDT, calculated as a function of the beltline material chemical composition (copper and nickel contents) and the neutron fluence according to the procedure-given in the PTS to distinguish it from other procedures for calculating rule, and called RTPTS RT The greater the e. mounts of copper, nickel and neutron fluence the NDT.

f r the' material and the lower its fracture resistance.

The higher the RTNOT screening criterion is 270*F for plates and axial welds and 300*F for the

- ircumferential weld. The rule does not consider the effect of vessel c

operating temperature and material surveillance test results on the calculated using the RT The rule is currently being arnended to calculate the RTPTS PTS.

trend curves in Regulatory Guide (RG) 1.99, Rev. 2.

The licensee, in response to our concerns about embrittlement; provided the folicwing significant information:

_1.

The reported copper and nickel contents of the weld metal are now assumed to be higher,-be:ause the actual values are unknown, and the licensee elected to report measurements made for a " sister" vessel, the Belgian BR-3 reactor, instead of previously-reported measurements for a weld in the upper head of the Yankee Rowe vessel.

2.

-The nominal operating temperature is 500*F, whereas the data base for-R.G. 1.99, Rev. 2 and the PTS rule"is from reactors that operate at a nominal' temperature of 550'F.

(Lower irradiaticn temperature increases RTNDT*)

The surveillance data from the Yankee Rowe vessel, all of which date from 3.

the late-1960's, show high sensitivity to neutron embrittlement, even

l considering the effect of the lower irradiation temperature. These data were known to the AEC but were discounted because the operating tempera-ture in the f trst few fuel cycles was known to be low (500'F), and there were coast down periods involving low operating temperature of several months duration at the end of the fuel cycles.

11.2.2 ' Evaluation of Material Proporties The beltline in the Yankee Rowe reactor vessel consists of an upper plate, a lower plate, two axially oriented welds'and one circumferential1y oriented weld. The only surveillance data from these materials is from the YNPS beltline upper plate. -The chemical composition and heat numbers for the upper and lower plates are known.

The chemical composition and heat numbers for the axial and circumferential welds are unknown. Eighty-five percent of the accumulated irradiation occurred at a cold leg temperature between 500'F and I

520*F. The remaining fifteen percent of the accumulated irradiation occurred.

at cold leg temperatures less than 500'F.

f l

'The staff's estimate and licensee's estimate of the mean value reference temperature in 1990_for each Yankee Rowe beltline material at its peak neutron flux location are tabulated in Table I.

The mean value reference temperature is the sum of the unirradiated reference temperature and the increase in reference temperature resulting from neutron irradiation at an irradiation

-temperature of 500'F. The staff's estimate of the increase in reference i

l temperature was estimated for_the peak neutron fluence in-1990 at the inside 19 surface of.the reactor vessel.

The peak neutron fluence is 2.3 x 10 n/cm8 19 for the upper shell plate, 2.05 x 10 n/cm2 for the loter shell plate and circumferential welds, and.38 x 10 n/cm2 for the axial welds.

The neutron

- fluences were calculated by the licensee using a methodology documented in

. letters from G. papanic, Jr. dated January 22, 1986, October 28, 1986 and February 4, 1987. The staff review of the licensee neutron fluence calculation i

methodology is documented in a letter to the licensee dated March 10, 1987.

The licensee is currently recalculating these fluences.

The results of this analysis will not be available before October,1990.

w,

10

!!.2.2.1 Upper Plate The licensee's estimate of the increase in reference temperature for the upper plate was derived from Yankee Rowe and BR-3 surveillance data, but did not correct-the BR-3 data (irradiation temperature 525-540'F) to account for the In lower irradiation temperature (500'F) of the Yankee Rcwe reactor vessel.

addition, the licensee doubled the neutron fluence values reported for the Yankee Rowe veillance data. The licensee did not include the effect of i

lower irradiation temperature in its analysis because they claim that the coarse grain size of the upper plate surveillance material eliminates the effect of irradiation temperature.

The licensee's coarse grain theory is based on an argument that irradiation-inducad defects in a coarse grain structure are more stable than irradiation-induced defects in fine grain Since the irradiation-induced defects are more stable in the structures.

coarse grain structure, the licensee concludes that the lower irradiation temperature of its reactor vessel will not affect the BR-3 data Because of l

Very limited surveillance data applicable to the Yankee vessel, the staff does not consider that the licensee has yet substantiated this theory.

A literature survey performed by the staff revealed three reports which indict.te irradiation temperature has an effect on neutron irradiation In rieference 1 (Stallman, ORril), irradiation temperature was embrittlement.

found to increase transitica temperature by 0.5 to 1.5 degree per degree decrease in irradiation temperante from 550*F, for a heat of A 533-B plate (the 02 plate from the ORNL HSST program). Odette (Ref. 2) has similarly found a factor of 1 degree per degree using a large data base of surveillance data.

In addition, Lowe (Ref. 3) has found about 0.7 degree per degree change in irradiation temperature, for Linoe 80 welds. Overall, these factors are probably dependent on the composition, processing history, etc. of the steel.

Although, References 1 and 2 do not specifically address coarse grain structures, the staff included the irradiation temperature effect in its evaluation because the licensee has not presented any Charpy data that shows l

the reference temperature for its plate material does not increase with a t

O

A

i.. :

g g-decreasein? irradiation _ temperature. The staff estimate of the reference 4

temperature;includesLa correction for irradiation. temperature and is based on the analysis performed by Odette (Ref. 4).

II.2.2.2 Lower Plate

~

-The licensee's estimate of the' increase in reference temperature for the lower pir.te was derived from Yankee Rowe and BR-3 surveillance data, but was not corrected.for lower irradiation. temperature or the increase in the amount of~

nickel in the. lower plate compared to the amount in the surveillance plate.

lThe lower plate has 0.63 percent nickel and the surveillance plate has 0.18 percent nickel. The' licensee believes no correction is necessary because of the postulation _that the coarse grain of the plate eliminates the nickel and

+

irradiation temperature effects.

To support the conclusion that-the nickel effect may be eliminated for coarse. grain structural material, the licensee reports the conclusions df a 11aricchiols-(Ref. 5) study.- In-this study, " Nickel was reported to reduce the_ damage introduced by neutron irradiation up to a content of about 1.0

- This study appears to contradict the results from a statistical percent."

9e results of the analysistof_ commercialljS reactor surveillance data.

statistical analysis of base meta 1' surveillance data is reported Table 2 of RG L

zl.99, Rev.2, which is conta ned here as Table-2.

This Table indicates that for a i

particular amount of copper, nickel: increases the chemistry factor, which results

'in=anincreaseinthematerial'sreferencetemperature(damage), nota-decrease i

as reported in the Maricchiols study.

Since the statistical analysis performed to derive.the-chemistry factor in the tables in RG 1.99, Rev. 2 indicates that there is a nickel effect and the licensee has not provided any' data from_ coarse grain-structure material that shows there is no nickel effect, the staff concludes there is a nickel effect.

The staff estimates that an increase in nickel from.18 percent to.63 percent at 500*F irradiation temperature results in an 80'F increase in the reference The staff Thjs;value is based on analysis by Odette (Ref. 4).

temperature.

i I.

r 12 con'siders that it is important in order to determine whether longer term

. operation should be authorized to determine the effect of coarse grain for operating' temperature and metal chemistry representative of the Yankee Rowe vessel.

!!.2.2.3 Circumferential and Axial Welds The'circumferential weld is one of the critical materials. The axial welds are not because they are exposed to only one-sixth of the peak fluence due to their azimuthal location relative to the core.

The licensee estimated the increase in reference temperature for the circumferential nelds using the methodology recomended in RG 1.99, Rev. 2 and a correction-facter for irradiation temperature.

As-discussed previously, the chemical composition of-the Yankee Rowe beltline welds is not known. The f.

licensee used the' chemical composition of a BR-3 weld to estimate the increase s

in reference temperature resulting from neutron irradiation. The licensee believes that the amounts of copper (.183 percent) and nickel (.70 percent),

reported for the BR-3 weld may be used as estimates for their welds because the BR-3 weld and Yankea Rowe beltline welds were fabricated by the same vendor Babcock Wilcox, using the same process (submerged arc) and the same l procedures =(copper-platad filler wire with Linde 80 flux). However, this conclusion'is not supported by industrial experience. The B&W Owners Group (Ref. 6) evaluated the weld chemistry of Babcock & Wilcox fabricated Linde 80

-w= 4s.

The reports indicates that the total copper concentration in the weld metal results from a combination of the amount of copper plating and the base filler wire alloy concentration.

However, the principle source of copper in the as deposited weld metal is the amount of copper plate. Reference 6 indicates the amount of copper varies from heat.of wire to heat of wire.

L Until the licensee. determines the chemical composition of the circumferential and axial' welds, the amount of coppv in the welds should be considered unknown g

L and bounding values of copper should be used to estimate the effect of neutron irradiation on the weld metal's reference temperature, f

4 13 The d&ff used two bases for estimating RT for the circumferential weld.

NDT One method uses a set of data compiled by Odette (Ref. 4) for 500*F irradia-The other method uses RG 1.99, tion, which yields a 370*F value for RTNOT.

Rev.-2 methodology,-bounding values for copper and nickel, 0.35 percent and l

0.70 percent respectively, and 50'F for the irradiation temperature effect.

Figure 1 (Figure 4 from Reference

- This yields a value of 330'F for RTNDT.

1) reports the increase in reference temperature for weld metals and base metals (plates) at irradiation temperature of 500'F. The dashed line has been added to represent the increase in reference temperature for the circumferential weld using the RG 1.99, Rev._2 bounding method with 50'F correction for the irradiation temperature effect.

Since this curve bounds all the existing weld data in the Odette report, this method has been used to estimate values of of reference temperatures for the circumferential and axial weld metal where the amount of copper is unknown and the weld metal is subject to 500*F irradiation temperature.

The predicted value of the reference temperatures in 1990 for the circum-ferential weld and longitudinal welds are 330'F and 226'F, respectively.

These values are for high copper welds.

If the chemical analyses of these welds indicates that the amounts of copper are significantly less than 0.35 percent copper and 0.70 percent nickel, the reference temperatures will be significantly reduced.

For example, if the circumferential weld had 0.20

- percent copper and 0.70 percent nickel, the reference temperature would be 262'F (212*F from RG 1.99, Rev. 2 and 50'F for irradiation temperature effect).

Thus, the staff considers that it is important in order to determine whether longer term operation should be authorized to determine the actual chemical composition of the circumferential weld.

11.2.3 Sumary values for Yankee The level of uncertainty is higher for the estimates of RTNDT Rowe than has been encountered for other reactor vessels. Therefore, con-sidering the uncertainty in weld chemistry and the effects of coarse grain, the staff believes the RT f r both the lower plate and the circumferential weld NDT should be assumed to be 350*F : 50*F.

m

14 11.2.4 Probabilistic Fracture Mechantes Although the Yankee Rowe reactor vessel beltline has not received any inservica volumetric inspection, other areas of the reactor vessel have been inspected.

These inspections re,' ort that the welds do not contain any flaws exceeding the acceptance limits defined by 10 CFR 50.554 and ASME Code Section XI.

In developing the PTS rule, the staff used a " Marshall" distribution (Ref. 7) of flaws. The " Marshall" distribution, which was developed in the mid-seventies, characterized deftets in a vessel entering service, including defects considered acceptable according to fabricatien codes and undetected during inspection.

The Yankee Rowe reactor vessel beltline was fabriacted using methods and materials similar to othe, commercially operated reactor vessels except that the clad in the Yankee Rowe reactor vessel is spot-welded and the clad in all, other concercially operated reactor vessels is fusion welded.

Hence, except for the effect of spot welding, the distribution of flaws in the Yankee Rowe reactor vessel should be similar to the distribution in other cocnerically operat.c reactor vessels.

During th? Summer 1990 refueling outage, the licensee ultrasonically examined the reactnr pessure vessel closure head and upper regions of the pressurizer, which containet. spot-welded clad similar to the clad in the reactor vessel beltline. The taff inspector (Ref. 8) concurred with the licensee's evalua-tion of the uit'asonic data that there was no extension of previously observed cladding cracks into the base metal. This inspection supports the conclusion that postulated cracks in the spot weld in the reactor vessel beltline cladding would not progress into the base metal due to the operation of the reactor vessel and the "Harshall" distribution appears to be applicable for the Yankee Rowe reactor vessel beltline.

However, until the licensee performs an intervice inspection of the beltline materials, the conditional failure prob-ability should be increased to account for the uncertainty in service-induced

flaws, i

4 15 To assess the effect of cracks on the probability of failure given the occurrence of a transient event, the licensee utilized probabilistic fracture mechanics analysis The staff gui6anc, for estimating the conditional probability of reactor vessel f ailure is provided in Regulatory Guide 1.15a.

Thermal and stress analyses for the vessel wall have to be perfortred.

Input for this analysis includes the primary system pressure, the temperature of the coolant in the reactor vessel downcomer, the fluid-film heat transfer ccefficient adjacent to the vessel wall, all as a function of time, and the vessel properties.

Probability density distribution functions for flaw size, crack initiation fracture toughness, crack arrest fracture toughness, and either the vessel materials nil. ductility reference temperature, or the vessel materials copper and nickel contents, and fast neutron fluence have to be developed.

For each transient of interest, many deterministic fracture mechanics analyses have to be performed to determine the number of times the crack penetrates through the vessel wall per 100,000 runs (forexample)asa, result of the stress level, flaw size, toughness and other variables selected for each run.

The calculations are performed with a probabilistic fracture mechanics computer code based on the Monte Carlo simulation technique.

The licensee has performed a probabilistic fracture mechanics analyses for several transients.

For example, the licensee performed a sensitivity study that predicts conditional probability of reactor pressure vessel failure is approximately 10~3 given the occurrence of a 1.3 inch-diameter small break LOCA event, which they believe is the controlling event, and for the reference temperatures reported in Table 3.

The reference temperatures used by the licensee are similar to the values estimated by the staff except for the lower plate.

The c1nditional failure probability for a small break LOCA event for the lower plate with a reference temperature of 325'F is less that 10-5 This piste has a low conditional failure probability at these high reference tem-peratures because only a small portion of the plate is in the beltline region.

Considering the results from the 325'F reference temperature analysis, a mean value of 355'F should not significantly change the conditional failures probability.

o O

l

16 When evaluating the results of the licensees sensitivity study one must cen-sider the assumptions used in the analysis. The licensee assumed a " Marshall" distribution of flaws and that cracks would arrest according to the average crack arrest data (Ref. 9).

The flaw density distribution function used by the licensee may not be representative of the Yankee Rowe reactor vessel because of its unique spot cladding on the inside surface of the reactor vessel.

It also appears that the licensee's analysis may not have adequately acccunted for the low upper-shelf energy of the vessel material which affects the " arrest" of initiated cracks. Given these apparent deficiencies and others that have been noted to date, the staff does not accept the licensee's estimate of the condi-tional failure probability of the reactor pressure vessel.

The staff and its contractor are continuing a detailed review of the licensee's analysis.

The The review of this analysis should be completed by the end of October 1990.

results of this review will be important in determining future action in connection with this license.

In view of these uncertainties the staff is unwilling to accept the licensee estimate of coi3ttional vessel failure prob.

ability of 1x10~3 given a specific size small break LOCA.

In the meantime the staff judges it would be prudent to assume the conditional probability of reactor pressure vessel's failure to be in the range of 10'I to 10-2,

!!.2.5 PTS Conclusions As discussed above, the staff concludes that there are substantial uncertain-ties associated with weld chemistry and the effects of coarse grain plate material on the shift in reference temperature.

These uncertainties could result in reference temperatures significantly higher than the screening criteria specified in the regulations.

Recognizing these uncertainties, the staff concluded that a more conservative range of conditional failure prob-ability (by a factor of 10 to 100 relative to the licensee's estimate) was appropriate. -This range when coupled with estimates of likelihood of the occurrence of PTS events and consideration of the plant specific features at Yankee Rowe import 6nt to such events, leads the staff to conclude that opera-tion until the end of fuel Cycle 21 is acceptable from PTS considerations.

However, additional information to resolve these concerns is needed to determine whether to authorize longer term operation.

- - - ~ - - - _ _ _ _ _ _ _ _ _ _ - -

17 LOW TEMPERATURE OVERPRESSURIZATION (LTOP)

!!!.1 Systems Evaluation In addition to the PTS events described above, another class of transients that could induce fracture in a brittle reactor vessel beltline are low tem.

peratureoverpressure(LTOP) events.

These events could occur during plant heatup when pumps are being started and there are possibilities for the mis.

alignment of valves and ccntrols following maintenance operativ,s.

The occurrence of such events has led to requirements comprising a low setpoint relief valve and control circuitry as described in NUREG/CR-5186, (Ref. 10).

For LTOP-considerations analyses are divided into two general categories:

(1) mass (water) addition events and (2) energy addition e'.ents.

In its July 5

-1990 submittal the licensee presented analyses of such events for the Yankee Rowe plant. The analyses were based upon industry wide historical data on LTOP events from 1980-1986 adjusted by consideration of Yankee Rowe specific

• features.

The licensee concluded that the like11hoort of vessel challenges from

. LTOP events was very low.

The staff review in this area emphasized the applicability of historical data to Yankee, impact of Yankee specific LTOP system features; and administrative

.centrols used to minimize human errors, 111.2 LTOP Event Frequency

.For LTOP analyses the licensee used the method and data described in NUREG/

CR-5186(Ref.10).

Feu Jres important for Yankee relative to the generic data base are:

Feature A:

The RHR (Shutdown Cooling System) at YNPS is a dedicated system which is different from most plants. The system is connected to the Main Coolant system through dual isolation valves. The suction to the Shutdown Cooling pump is from the H cold leg loop. There are two pumps and heat

~-

4 l

18 exchangers for redundancy. There is also a relief valve on both the suction and return lines for overpressure protection.

Feature B:

The FORY (in the low setpoint condition) and the shutdown cooling relief valves are required to be operable by Technical Specifications whenever the plant is in the Modes 4 and 5 and the system temperature is less than 300'F.

The shutdown cooling relief valves are tested when the plant is operating in Mede 1 and the shutdown cooling system is required to be isolated.

The PORY is tested when the plant is in Mode 6 with the reactor be d remov',J.

Feature C Plant procedures require that power be removed by lochng out the breakers for the Main Coolant pumps and the Safety injection pumps prior to being in a water solid condition.

power is reraoved from $1 pumps below 200'F.

Feature 0 The safety relief valves of the shutdown cooling system cannot be f,

automatically isolated once the system is placed into operation because t,he system isolation valves do not have any automatic isolation capabliity.

Feature Et During water solid condition operations, a dedicated operator is stationed to prevent or terminate any pressure excursion.

During operation below 300'F, 2 shutdown cooling relief valves and 1 p0RV are available to mitigate LTOP events.

In this temperature range, and with no credit for human intervention during an event, the licensee estimate of vessel challengoeventfr'quency(eventswheremitigationsystemsfail)is6.3X105 per reactor year.

NUREG 5186 reports a frequency of 2.5x10~3 per reactor year using generic data.

The difference is attributable to 2-factorst (1) the availability of an additional relief path at Yankee relative to generic data assumptions; and (2) a power lockout requirement for MCP and $1 pumps at Rowe l

which precludes energy addition events such.as were reported in the generic data-base.

l The staff-judges that the specific features of Yankee Rowe would reduce the likelihood of the vessel challenges from LTOP events in the operating range when the PORY is reset to the lower setpoint and the SDC system SRVs are l

19 available. An event frequency of 1x10*3 per reactor year was therefore chosen as a conservative screening value to assess the importance of LTOP events in this temperature r uje relative to PTS events.

Between 300'F and 330'F the SDC system is isolated, and above 380'F and 450 psig the PORV is reset to 2500 psig.

For all temperatures greater than 180'F a pressurizer bubble is required.

In the range of 300'F to 450'F a dedicated operator is required whose only responsibility is LTOP protection (by maintaining a 400 psi margin to the Appendix G curve).

Power is also removed from 2 of 3 safety injection purps at these conditions and all SI pump switches must be in pull to lock.

Inadvertent SI (which could cause a maximum pressure of 1550 psig) would therefore require a spurious $1 signal plus failure to have the $1 pumps in pull to lock.

In addition, the auto safety injection signal is blocked until 1800 psig.

The licensee concluded that the most probable LTOP challenge in this range (T greater than 300'F) is a charging / letdown mismatch.

A charging / letdown mismatch involving all 3 pumps could allow 100 gpm injec.

tion.

This rate would allow 10 minutes for operator action to preclude Viola-tien of the Appendix G curve in the event of a PORY failure to open.

However, even without credit for operator action, the licensee's frequency estimate for an event that would challenge the vessel is about 1x10-5 per reactor year.

This estimate assumes a PORY failure rate of about 10'I per demand, a mismatch frequency of 10-2 per reactor year, and the fraction of time the plant would be operating in the temperature range per year (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in 600 shutdown hours) or 10-2 per reactor year.

In view of the licensee's analysis and the historical data regarding challenges to systems with a pressurizer bubble and FORV (zero events), the staff considers that the screening value of 1x10'3 per reactor year discussed above for LTOP below 300'F is also conservative in the temperature range above 300'F.

Above 380'F and 450 psig the PORY is reset to 2500 psig.

However in this range the vessel temperature is high enough that brittle fracture is of negligible concern.

20 f

!!!.3 LTOP Materials Evaluation The licensee did not discuss materials aspects of LTOP events in their reports.

The staff calculated the conditional probability of vessel fracture based on the peak pressure for the Yankee Rowe vessel using the methods set forth in Reference 11 and assuming RT is 320'F.

An LTOP peak pressure in the range NDT 1000-2000 psig has a conditional probability of vessel fracture in the range 10~3 to 10-2, 111.4 LTOP Conclusion Based upon a conservative screening value of 1x10~3 per reactor year for LTOP event frequency and a conditional vessel failure probability for LTOP events of 10-2 to 10~3, the staff concludes that PTS events are bounding for brittle fracture considerations.

!Y.

UPPER-SHELF EllERGY EVALUATION IV.1

Background

Reactor vessel beltline materials are required by Appendix G to 10 CFR Part 50 to have adequate fracture toughness.

Specifically, beltline materials are required to have Charpy upper-shelf energy (USE) no less than 50 ft/lb throughout the life of the vessel. Otherwise, an analysis, approved by the staff, to demonstrate the existence of margins of safety against fracture equivalent to those of Appendix G of the ASME Code is required.

IV.2 Upper-Shelf Energy Events - Mr+erial Evaluation in a letter dated May 1, 1990, the staf# informed t5e licensee of the results of analyses that indicate that the USE for the Yankee Rowe vessel could be as low as 35.5 ft/lb.

The staff specified the regulatory requirements that had to be met for vessels with USE below 50 ft/lb and provided the USE evaluation criteria based on current developments of the ASME Code.

At present, these l

criteria have only been developed for ASME Code Service Levels A and B, e.g.,

l

21 Normal and Upset loading conditions.

The staff believes that Service Level C and D, i.e., Emergency and Faulted conditions, criteria are unnecessary because, except for PTS and ATWS transients, Service Level C and 0 loads do not exceed level A and B loads.

PTS events are discussed above. With regard to ATdS, the staff reviewed results of ATWS analyses which the licensee has submitted in 1974.

The peak pressure estimated for a loss of feedwater ATWS was estimated to be 2820 psig.

Since the licensee's Charpy USE analysis assumed an RCS pressure of 3437 psig the staff concludes that ATWS events are reasonably bounded by the licensees USE analyses.

The licensee performed an USE analysis for Normal and Upset loading conditions, i.e., ASME Code Service Levels A and B, using the ASME Code criteria now in preparation.

The ASME code criteria now in preparation will require margins of safety against fracture equivalent to those required by the regulations.

Based on a preliminary review of the licensee's analysis, it appears that the licensee's analysis satisfies the ASME code criteria for Service Levels A and B and provide margins of safety against fracture equivalent to those required by Appendir G of the ASME Code.

The licensee also performed a low USE analysis for two of the PTS transients. The effects of low USE on crack arrest will l

also be considered in the PTS analysis being evaluated by the staff's contractor.

V.

CONCLUSION In order to address several NRC concerns with respect to the requirement for reactor vessel fracture toughness for protection against pressurized thermal shock events, the Yankee Rowe licensee has provided an analysis of the potential events leading to a challenge to the reactor vessel.

That analysis addressed both the probability of the initiating events as well as the probability of a pre-existing crack propagating through the vessel wall.

The l

licensee also estimated the likelihood of challenges to the vessel from low temperature overpressurization events.

As discussed above, there are a number of areas in which the staff concludes that additional safety margin or con-servatism in the analysis would be appropriate; and that additional informatien I

O

t i

22 to fully resolve the areas of concern is f.eeded in order to determine whether longer term operation should be authorized.

Actions required of the licensee during the next operating cycle are specified below.

However, in the interim, the staff concludes that reasonable assurance of the public health and safety is provided since the potential for reactor vessel f ailure is very unlikely.

VI.

FUTURE ACTIONS In order for the licensee to demonstrate that longer term operation can be carried out without undue risk to the public health and safety, the licensee should provide the NRC, within 60 days af ter restart, a detailed plan of action.

The following elements should be included in the plan:

VI.1 Short Term (Completed within 3 months) 1.

Peer review of YAEC 1735, " Reactor Pressure Vessel Evaluation Report forr Yankee Nuclear Power Station."

2.

Revise fluence calculations.

VI.2 Long Term (Completed prior to Cycle 22 startup) 1.

Develop inspection methods for the beltline welds and each beltline plate from the clad to 1 inch from the clad / steel interface to determine if the metal contains flaws, 2.

Perform tests on typical Yankee Rowe base metal (0.18-0.201 Cu) to i

19 2

determine the effect of irradiatioa (f a 1-5X10 n/cm ), austenitizing temperature (1650'F-1800'F)andnickelcomposition(0.18-0.70 percent)on l

embrittlement at 500'F and 550*F irradiation temperatures.

3.

Determine composition of the circumferential weld metal in beltline by removing samples from the weld.

23 In addition, the licensee should install surveillance capsules in accelerated irradiation positions. The capsules are to include materials representing the beltline circumferential weld metal and upper and lower plates.

Yll.

REFERENCES 1.

F. W. Sta11mann, " Curve Fitting and Uncertainty Analysis of Charpy Impact Data " USNRC NUREG/CR-2408, January 1982.

2.

G. R. Odette and G. E. Lucas, " Irradiation Embrittlement of LWR Pressure Vessel Steels," EPRI NP 6114, January 1989.

3.

A. L. Lowe, "An Evaluation of Linde 80 Submerged Arc Weld Metal Charpy Data Irradiated in the HSST Program," ASTM STP-1046 Vol. 2, 1990.

4.

G. Robert Odette, Acting Dean, College of Engineering, UCSB "1990 shift.

Estimates for The Yankee Rowe Vessel," July 30, 1990.

5.

Maricchiolo, C., Milella, P. P., and Pini, A., " Prediction of Reference Transition Temperature Increase Due to Neutron Irradiation Exposure;"

Radiation Erbrittlement of Nuclear Reactor Pressure Vessel Steels:

An InternationalReview(SecondVolume),ASTMSTP-909,L.E.Steele,Ed.,

American Society for Testing and Materials, Philadelphia, 1986, Pages96-105.

6.

B&W Owners Group Report BAW-1799, "B&W 177-FA Raattor Vessel Beltline Weld Chemistry Study," July 1983.

7.

W. Marshall, An Assessment of the Integrity of PWR Pressure Vessels, United Kingdom Atomic Energy Authority, October 1976.

8.

Letter from H. Kaplan and J. O'Neil, " Yankee Rowe Feeder-Ultrasonic Examination of Pressurizer and Reactor Vessel," August 15, 1990.

24 e

1 i

9.

F. A. Simonen, et al., " VISA.!! - A Computer Code for Predicting the Probability of Reactor Vessel failure," Battelle Pacific Northwest Laboratories, USNRC Report NUREG/CR-4486, April 1986.

s i

10.

B. F. Gore, et al., PNL, "Value-impact Analysis of Generic issue 94,

" Additional Low Temperature Overpressure Protection for Light Water 8!

Reactors," NUREG/CR-5186, November 1988 af 11.

C. Y. Cheng, Chief. EMCD memorandum to Robert C. Jones, Chief SRXB,

'[

" Conditional Probability of Vessel Fracture from LTOP Events," August 9, i

1990.

12.

FAX, Jane Grant, Yankee, to Pat Sears, NRR, 8/8/90 - Material Properties Answers to Questior.s at 8/7/90 Meeting.

13.

FAX,' Jane Grant to Pat Sears 8/10/90 - Answers to Questions at 8/7/90 Meeting.

14.. FAX, Jane Grant, Yankee, to Pat Sears, NRR, 8/10/90-- Answers to l

Questions at 8/7/90 Meeting.

15.

FAX, Jane Grant, Yankee, to Pat Sears, NRR, 8/10/90 - Answers to i

Questions at 8/7/90 Meeting.

l I

L 16.

FAX, Jane Grant, Yankee, to Pat Sears, NRR, 8/14/90 - Answers to Questions at 8/7/90 Meeting -

17.- FAX, Jane.Crant, Yankee, to Pat Sears, NRR, 8/14/90 - Answers to Questions at 8/7/90 Meeting..

18. -Yankee letter dated 8/3/90 - Pear Review of Reactor Pressure Yessel L

Evaluation.

1

19. -Yankee letter dated 8/2/90 - PTS Sensitivity Study.

v me-o-

.v-

...-v ve n u

e a.-

e-mm-ow.-,=-m.e-w,,

eyew.rp+rra.--w,--n+-o

• -, e n w -re.rw ww m *.s r e e,-eea+.ee.n.rs t

ee w

e, r

r

-wm*mr-e

25 j

20.

FAX, Jane Grant, Yankee, to Pat Sears, NRR, 8/14/90 - Updated Table 5.7 of 7/5/90 submittal.

21.

FAX, Jane Grant, Yantee, to Pat Sears HRR, 8/17/90 - Answers to Questions by G. Kelly, NRR at 8/16/90 Telecon.

22. FAX, Jane Grant, Yankee to Pat Sears, NRR, B/17/90 Answers to questions by G. Kelly, NRR at 8/16/90 Telecon.

23. FAX, Jane Grant, Yankee, to Pat Sears, NRR, B/27/90 - Fracture Mechanics Results.

4 9

e e

i

TABLE I LICEllSEE AND STAFF ESTIMATES OF REFERENCE TEMPERATURE.

RT FOR THE YNPS BELTLINE MATERIALS IN 1990 NOT YNPS-Increase-in Ref.

Beltline-Unitradiated Temp. Resulting.

Ref. Temp.,

Material Ref. Temp.('F) fromtrrad.(*F)

RT in1990('F)

NDT Staff Licensee Staff Licensee Staff Licensee.

Estimate Estimate Estimate Estimate Estimate Estimate' Upper Plate 30 10 245 180 275 190 Lower Plate 30 10 325 173 355 183 Axial Welds 10 10 216 131 226 141 Circum.

farential Weld 10 10 320 219 330 370 229

-O

~

. _ _ - -. _... _.. -....... _.... -, _ _... _ _ _.. -..,, _ _ _ _ - _,., _, _ _ _ _. - _... -.. -., _.,. _,....... ~ ~. _ _. _ - _ _

e TABLE 2 CHEMISTRY FACTOR FOR BA$E METAL., 'F CeTt.

Ns(kel. Wi 9

~

W 0

0 20 0 40 0 60 0 60 t 00 t ;o 0

20

D 20 20 0

D o

0 01 0

0 20 00 20 20 I'o 0 02 l0 20 20 20 20 0

o 0 03 20 20

D 20

o

0

o 0 04 22 26

, 26 26 26

6 6

0 03 23 31 31 31 31 31 31 0 06 28 37 37 37 17 37

)7 0 07 31 43 44 4

4 4

u 0 08 4

48 51

$1 St

$1 31 0.09 37 33 38 38 38 38 33 0.10 di 38 65 63 67 67 67

0. l i 45 62 72 74 77 77 17 0.12 49 67 79 83 66 86 16 0,13 33 71 91 96 96 96 t

0.14 37 75 91 100 10$

106 106 -

0.13 al 80 99 110 113 117 ll?

O.16 65 84 104 lit 123 125 125 0,17 69 18 110 127 132 135 135 0.18 73 92 113 134 141 lu 144 0.19 78 97 820 142 ISO 134,

154 0.20 82 102

• 123 149 139 164

, 165 0 21 86 107 129 135 167 172 I?4 1

0.22 91 112 134 161 176 181 154 0.23 95 ll?

138 167 184 190 194 0.24 100 121 143 172 198 199 204 0.23 104 126 148

!?6 199 208 214 l

0 26 109 130 151 180 20$

216 221 l

0.27 114 134 ISS 164 211 223 230 0.28 119 138 160

!$7 216 233 239 0.29 124 142 IM 191 22l 24l 248 0=30 129 144 167 194 223 249 257 0.31 l'~

l.11 172 198 228 235 266 0.32 135 173 202 231 260 274 i

0.33 14 160 180 20$

234 2M 282 0.34

.6 IM IM 209 238 268 290 0,35 133 IH 187 212 241 272 298 0 36 138 173 191 216 245 275 303 l

-0.37 162 177 196 220 248 278 308 0.38 It4 182 200 223 250 281 313 0.39 171 185 203 227 254

- 285 317 0.40 173 189 207 231 237 288 320 1

l l

l l.-

7

.. i

i 4

l TABLE 3 COMPARIS0N OF REFERENCE TEMPERATURES ESTIMATED BY,THE STAFF AND VALUES USED BY THE LICENSEE IN ITS SENSITIVITY STUDY l

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UNITED STATES e

J7 i NUCLEAR REGULATORY COMMISSION 4

WAEHINGTON. D C. X6fi6 June 25, 1991 Docket No.50-029 (10 CFR Section 2.206)

Diane Curran, Esq.

Harmon, Curran, Gallagher & Spielberg 2001 S Street, N.W.

Suite 430 Washington, D.C. 20009-1125

Dear Ms. Curran:

I am writing to acknowledge receipt of the " Petition for Emergency Enforce-ment Action and Request for Public Hearing" (Petition) submitted by you on behalf of the Union of Concerned Scientists and the New England Coalition on Nuclear Pollutien (Petitioners). On June 4, 1991, the Petition was submitted directly to the Commissioners of the U. 5. Nuclear Regulatory Conmission (NRC). The Petition was filed in accordance with Section 2.206 of Title 10 of the Code of federal Regulations (10 CFR 2.206) and thus should have been filed with the Executive Director for Operations.

However, the Petitioners seek relief directly from the Commissioners because they believe that the NRC staff has failed to properly execute its responsibilities in this matter in permitting the Yankee Rowe Nuclear Power Station to continue to operate through Cycle 21 (approximately February 1992).

The Petition has been referred to me for treatment under 10 CFR 2.206.

The Petition seeks the immediate shutdown of the Yankee Rowe facility of the Yankee Atomic Electric Company (licensee) based upon allegations that the Yankee Rowe facility is operating in violation of HRC requirements for reactor pressure vessel integrity and that the NRC staff's Safety Assessment of August 30, 1990, contains a number of deficiencies.

The Petitioners argue these reasons prove that the continued operation of the Yankee Rowe f acility poses a serious threat to public health and safety.

The Petitioners further request that the Yankee Rowe facility remain shut down until it complies with regulatory requirements and that the Commission provide a sublic hearing, with rights of discovery and cross-examination, to determine tle regulatory compliance before permitting the facility to resume operation.

The Petitioners allege specifically that the Yankee Rowe reactor pressure vessel failed to meet NRC requirements. The Petitioners argue that the Yankee Rowe facility does not comply with the requirements in 10 CFR 50.61 regarding reference temperature for reactor vessel material, the requirements in Appendix G to 10 CFR Part 50 regarding fracture toughness and the require-i ments in Appendix H to 10 CFR Part 50 regart:ng a surveillance program for reactor vessel material.

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i Diane Curran, Esq.

?-

On August 31, 1990, the NRC staff issued its " Safety Assessment of Yankee Powe Yessel" (Safety Assessment) concluding that the Yankee Rowe facility could operate safely through Cycle 21. The petitioners make specific allegations that the NRC staff's Safety Assessm e t is deficient. The Petitioners argue that the Safety Assessment contains eirors and insufficient information in the assurrptions underlying the calculations regarding the amount of neutron irradiation absorbed by the reactor vessel, the temperature of the metal during the time it is exposed to neutron irradiation and the chemical corrposition of the metal.

in addition, the Petitioners argue that the Safety Assesstnent is inconsistent with the NRC policy on Safety Goals and that it failed to take into account the explicit recornmendation of an NRC staff expert on reactor pressure vessel integrity that the Yankee Rowe f acility not be permitted to operate.

The Petition presents no new information in regard to the integrity of the reactor vessel at the Yankee Powe facility. The Petition cypresses disagree-ment with tht f20 stafs conclusions reached in the Safety Assessment that the Yankee Powe f acility was safe to operate through Cycle 21. The NRC staff has reviewed the Petition and has found no new information that would call into question the conclusions reached in its Safety Assessment.

In making the Safety Assessment, the staff considered the views of NFC staff expert Dr. Parda11 as did the Advisory Conoittee on Reactor Safeguards (ACRS) which reported f avorably regarding continued operation of the Yankee Powe f acility.

(See letter of Septenber 12, 1990, from ACRS, inclosure 1.) The assertion that continued operation of Yankee Powe constitutes a serious threat to the public health and safety because of the six alleged violations of NRC requirements is without trerit for the following reasons.

The Petition ir.dic6tes that the reference temperatures for the upper plate, the lower plate, and the circumferential wtid exceed the screening criteria for pressurized thermal shock (PTS) in 10 CFR 50.61(b)(2). The licensee, as documented in Report YAEC t:0, 1735, July 1990, reports that the reference terrperatures are below the PTS screening criterion. However, the NRC staff believes the PTS screening criterion may have been exceeded.

That belief is based on conservatively considering the uncertainties associated with weld chemistry, irradiation temperature, grain size effects and flaw distribution as noted in the EC staff Safety Assessment transmitted to the licensee by the letter of August 31, 1990.

10;CFC 50.f1 does not require shutdown if the FTS screenir.g criterion is exceeded 7 The NPC may, as specified in 10 CFC 50.61 (b)(5), on a case-by-case basis, approve operation of the facility at values of reference temperatures in excess of the PTS screening criterion. The rule requires the staff consider factors significantly affecting the potential for failure of the reactor vessel including the results of a probabilistic fracture mechanics analysis in reaching a decision to approve operation.

The NRC staff also believes that the reference temperatures for axial welds in the upper and lower plates nay also exceed the PTS screening criteria, as indicated in the October 9,1990,memorandumtoACRS(Enclosure 2). The NRC staff requested L

the licensee to perform a probabilistic fracture mechanics analysis using l

conservative values of reference temperatures for PTS specified by the l

Diane Curran, Esq.

-3 staff.

Those reference temperatures exceed the screening criterion. The NRC staf f reviewed the results f rom the probabilistic f racture mechanics analysis and considered the uncertainties resulting from low upper shelf energy (USE) of the vessel materials, the lack of beltline inspection, and the reactor vessel's unique spot-welded cladding. Therefore the NRC staff judged it to be prudenttoassunetheconditionalprobabilityofreactorpressurevesselfailure to be ir the range of 10E 4 to ICE-2 and the estimated frequency of the limiting PTS transient to be ICE-3 per reactor year.

Based on this assessrent the NRC staff authorized the licensee to operate the Yankee Rowe reactor vessel until the end of fuel Cycle 21.

The Petition indicates that the Yankee Rowe vessel uppcr plate is below the regulatory requirements for Charpy USE in 1: CFR Part 50, Appendix G Section IV.A.I.

However, Appendix G also indicates (1) that reactor vessels may be operated at lower values of Charpy USE, if operation is approved by the Director. Office of tiuclear Reactor regulation, and (2) that lower values of Charpy USE provide rargins of saf ety against f racture that are equivalent to those recuired by Appendix G of the ASPE : ode.

The licensee provided a fracture mechanics analysis in Report YAEC No. 1735, July 1990 to demonstrate thattreYankeeRowereactorvesselwouldhaveeauivalentmargInsoffracture toughness to those required by Appendix G of the ASME Code with a Charpy USE of 35 foot-pourds, in its August 31, 1990, Safety Assessments, the NRC staff reviewed the licensee's aralysis and approved the operation of the Yankee Powe reactor vessel at levels of Charpy USE less than the limits in Section IV. A.I. of Artendix G to 10 CFR Part !O.

The Petition alleges that the !;EC staff's Safety Assessment did not corsider the revised neutron fluence estimates, the vessel operating temperature, and the vestc1 composition.

The licensee revised its neutron fluence estimates in letters of September 28, 1990, and February 20, 1991. The NRC staff evaluated tR neutron 'luence estimates reported in the September 28, 1990, letter and docurcrted its findings in the October 9 1990, memorandum to ACRS (Enclosure 2). The peak rautron fluence estimates for the end of the current cycle (Cycle 21) reported in the licensee's February 20, 1991, letter are less than the values reported ir its September 28, 1990, letter.

Hence, the conclusions in the October 9,1990, UFC staff r,enorandum apply to the end of the current fuel cy cle.

In preparing the Safety Assesstent of August 31,199C, the NRC staff considered the effect of the reactor operating temperature and the uncertainty in vessel composition that are discussed in the Petition.

To account for the low operating temperature and the uncertainty in the vessel composition, the NRC staff increased the reference temperatures for the raterials.

These reference temperatures were evaluated as discussed herein.

The Petition elleges that the NRC staff did not consider that the beltline weld had not been inspected, had not received fracture toughness data from the licensee, and had not reviewed the licensee's analysis.

The NRC :.taff was aware that the licensee has not volumetrically examined the beltline welds in the Yankee Rowe reactor vessel since the plant began operating.

Therefore, the NRC staff in its probabilistic risk assessment assured that flaws existed in the reactor vessel plates and beltline welds in order to account for the uncertainty

Diane Curran, Esq.

4.

resulting from the lack of volumetric examination, in Report YAEC No. 1735, July 1990, the licensee provided fracture toughness data and an analysis to demonstrate equivalent margins to Appendix G as noted herein. The NRC staff reviewed the data and analysis and provided its Safety Assessment in the letter of August 31, 1990.

The petition asserts that the NRC staff's decision to allow the licensee to continue to operate Yankee Rowe is " flatly inconsistent with the Comission's

'$afety Goal' policy that the risk of a severe accident should be kept to less than one chance in a million." The Petition indicates that this conclusion is based on the NRC staff's own calculation that the risk of pressure vessel rupture is between Ex10E-5 and 5x100-6 and is thus greater than the Comission's large release reactor years) guidance of 1x10E-6 per reactor year (that is, one in a million The Safety Goal is not, and was never intended to be, a measure of adequate protection of public health and safety.

Rather, the safety Goal is a higher level of safety that the Commission believes the industry should strive to achieve. The Connission's Policy Statement on Safety Goals states the following:

Current reguletery practices are believed to ensure that the basic statutory requirement, adequate protection of the public, is met.

The Policy statement further states the following:

This statement of NRC safety policy expresses the Commission's views on the level of risks to public health and safety that the industry should strive for in its nuclear power plants.

The NPC staff's dccisions regarding plar,t operation are based upon adequate protection of the public health and safety, not the Commission's Safety Goal Policy.

The Petition indicates that Yankee Rowe does not have a surveillance program as required by 10 CFR Part 50, Appendix H, and has not had its vessel ultrason-itally inspected. The licensee discussed its surveillance and ultrasonic inspection program in Report YAEC No.1735, July 1990. The licensee and the NRC staff used the data from the licensee's surveillance program to assess the integrity of the Yankee Rowe reactor vessel. The beltline welds in the Yankee Rowe reactor vessel were volumetrically examined by radiography as a part of its fabrication quality control. All flaws detected that exceeded the acceptance criteria were removed and repaired. Although the licensee has not ultrasonically examined the beltline welds since the plant has been in service, it has examined other similar welds and observed no unacceptable indications. However to account for the uncertainty that flaws might be present, the NRC staff, in its probabil-istic risk assessment, assumed that flaws existed in the reactor vessel plates and beltline welds.

The Petitioners make the legal argument that compliance with NRC requirements is necessary to ensure that the Yankee Rowe facility operates safely.

However, the failure to comply with a particular NRC requirement does not necessarily mean that there is no longer reasonable assurance of adequate l

e Diane Curran, Esq.

June 25. 1991 protection of the public health and safety, particularly when the t4RC staff has evaluated the crea of alleged noncompliance and found that it does not pose an undue risk to the public health and safety.

The f4RC staff has tvaluated the Yankee Rowe reactor vessel issues carefully and has concluded that the vessel condition continues to provide adequate protection of the public health and

safety, in suneary, the Petitiontr's assertion that the alleged violations warrant irrediate action to shut down Yankee Rowe is without merit.

Accordingly, Petitioners request for emergency relief is denied. As required by 10 CFR 2.206, the NRC will address the specific issuts raised in the Petition within a reasonoble tine. is a copy of the tiotice that is being filed with the Office of the Federal Register for publication, Sincerely, Thomas E. Murley, Director Office of t:uclear Reactor Regulation Entlosures:

1.

Letter to K.

I',

Carr fn the ACRS 10/9/90

?.

tiemo to Committee on Reactor Safeguards, 10/09/90 3.

Related federal Register tiotico, cc: l'r. George Papanic, Jr.

Dr. Andrew C. Kadak

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ADVISORY COMMITTEE ON RE ACTOR SAFEGUARDS

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September 12, 1990 The Honorable Kenneth M. Carr Chairman U.S.

Nuclear Regulatory Commission Washington, D.C.

20555 Dear Chairman Carr

SUBJECT:

YANKEE ROWE REACTOR PRESSURE VESSEL INTEGRITY During the 365th meeting of the Advisory Committee on Reactor Safequards, September 6-7,

190, we discussed the degree and consequences of the Yankee Rowe reactor pressure vessel embrittlement due to neutron irradiation.

Our Subcommittee on Materials and Metallurgy discussed this matter with representatives of the NRC staff and the Yankee Atomic Electric Company during a meeting on September 5,

1990.

We also had the benefit of the documents referenced.

It has recently come to the staff's attention that the ref'erence temperature nil ductility transition (RTg7) of parts of the Yankee Rowe pressure vessel may substantially exceed the temperature limits for action delineated in the pressurized thermal shock (PTS) rule (10 CTR 50.61).

The main reason is that the Yankee Rowe core inlet temperature is about 50'F lower than that of other plants.

Another reason is the higher nickel content of the lower vessel plate.

These increase the rate of rise in RT with fast neutron g7 irradiation.

The exact value of RT for the vessel is uncertain because of:

g7 Uncertainty in the copper and nickel content of the e

circumferential weld near the reactor vessel beltline.

The absence of surveillance data for areas that appear to have e

the largest shift in RTg7, namely the circumferential weld,and the lower plate of the vessel.

Assurance of vessel integrity is further hindered by:

e The absence of any inservice inspection for flaws in the reactor vessel beltline region.

Such inspection has been infeasible due to the design of the vessel internals.

Relatively low toughness (low upper shelf energy) of the plate and welds near the core.

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The Honorable Kenneth M. Carr 2

Septembar 12, 1990

\\

Analysis of the various safety issues involved leads to the conclusion that PTS is the issue of most concern.

One bright spot in this picture is that several features of the plant's design make it less susceptible to overcooling events than more modern plants.

The licensee and the staff have both arrived at estimates of the shift in RT Both agree that the circumferential weld and the lower plate of.the pressure vessel have the highest RT g

in each case their estimates dif fer by about 150*F.

TN. However, licensee's representatives argue that due to the particular microstructure of the steel in the vessel, the shift in RT is independent of g7 irradiation temperature and nickel content.

We do not believe these arguments are

valid, and agree with the staff that temperature and nickel ef fects must be included in a valid estimate of the shift in RT An additional difference between the staff er.

and the licensee concerns estimates of the copper content of the circumferential veld.

There being no measurements for the composition of the circumferential weld and a large spread in copper values found in other plants, the staff prefers to choose a bounding value.

The applicant chose more of an average value.

In view of the uncertainty in the value for the Yankee Rowe vessel, we would choose the staff's bounding value.

Given that RT values for parts of the vessel probably exceed

(

g7 those requiring action under the PTS rule, is there significant risk in operating the plant?

The low probability of a PTS challenge leads to a low risk, even with a high RT Thus, we pT.

agree with the staff that operation for one more cycle is acceptable, provided the licensee initiate an active program to better characterize the material in the vessel near the reactor vessel beltline.

To do this the staff requires determination of the composition of the circumferential weld metal in the beltline l

by removing samples from the weld and development of an inspection method for the beltline welds and plate to depths of an inch below the inside surf ace of the vessel.

Both of these have been required by the staff for completion before the startup of the 22nd fuel cycle (now scheduled to begin in early 1992).

It is not clear that both can be achieved in that time, but certainly they should be l

accomplished in two fuel cycles.

l The staff also requires " tests on typical Yankee Rove base metal" to determine the effect of irradiation, austenitizing temperature and nickel content on embrittlement.

It is doubtful that any tests that the licensee could perform during the next fuel cycle would convince us that the effects of temperature and nickel on embrittlement are substantially different from those established by the much more extensive studies already available.

The effects are not well understood, and we believe prudence dictates tending l

more toward bounding values rather than best estimates based on j

limited new data that may become available.

The Honorable Kenneth M. Carr 3

September 12, 1990 l

However, the above will not adequately address the long-term operation of the plant.

This is the lead pWR plant in the industry's Plant Life Extension (PLEX)

program, and long-term operation with such large uncertainties in vessel integrity is unacceptable.

The extended operation of this plant would be acceptable only if e

A state-of-the-art ultrasonic inspection can be done on essentially all of the radiatior affected inner surface of reactor pressure vessel, e.g., one that complies with Appendices VII and VIII of Section XI of the ASME Code.

This inspection should also check for significant thinning !.n the lower head as a result of loose parts (irradiation capsules).

Continued operation would be dependent on the absence of significant

flaws, i

A reanalysis of the PTS question is made using well established

- e compositions for the material in the beltline region, or using limiting values of copper and nickel. This analysis should also include the fact that the crack. arresting ability of such material will be lower than more modern steel because of its low upper sheli sergy.

Such an analysis must show acceptable risk.

Sincerely, carlyle Michelson Chairman

References:

1. Letter dated July 5, 1990 from John D. Haseltine, Yankee Atomic Electric Company, to Richard Wessman, NRR, transmitting Reactor pressure Vessel Evaluation, dated July 9, 1990 2.

Letter-dated August 31, 1990 from Thomas E.

Murley, NRR, to Andrew C.

' Kadak, Yankee Atomic Electric Company, Subjects t

Yankee Rowe Reactor Vessel,.with Enclosure i

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NUCLEAR REGULATORY ('nMMISSION f

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A neswa.::::.. o c obst g,f OCT t 9 Iggo HEMORANDUM FOR:

Raymond F. Fraley, Executive Director Advisory Comittee on Reactor Safeguards t

FROM:

Thomas E. Murley, Director i

Office of Nuclear Reactor Regulation

SUBJECT:

YANKEE R0WE REACTOR VESSEL INTEGRITY Thank you for perr.iitting the NRC Staff to discuss with the ACRS Subcomittee on Materials and Metallurgy its assessment of neutron radiation embrittlement of the Yankee Rowe reactor vessel, in their letter to Chairman Carr, dated September 12, 1990, the ACRS agreed with the staff that even with a high RT operation for one more cycle is acceptable, provided the licensee inNa,tesanactiveprogramtobettercharacterizethematerialinthereactor vessel beltline.

such a program, The NRC staff and the licensees are discussing and planning in a letter dated September 28, 1990 (attached), the licensee reported preliminary results of its recalculation of the neutron fluence for the reactor vessel beltline.

The licensee reported that the neutron fluence at the peak flux location increased by 13 percent and the value at the minimum flux location increased b in neutron fluence, y approximately 300 percent. As a result of the increase the upper axial shell weld becomes the limiting mate' rial, with respect to PTS, and the egnditional failure probability cajculated "Si"9 the VISA 11 Code is 2.75 x 10".

This is within the range (10~ to 10~2) estimated by the staff in its safety assessment of the Yankee Rowe reactor vessel.

The licensee will submit final fluence values and a revised PTS analysis by November 30, 1990.

We will keep you informed on all integrity issues affecting the Yankee Rowe reactor vessel.

]

Thomas E. Murley, Director Office of Nuclear Reactor Regulation

Enclosure:

As stated l

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September 28, 1990 BYR 90-128 United States Nuclear Regulator / Commission Document Control Desk Washington, DC 20$55 Attention:

Mr. William Russell Associate Director for Inspection and Tedhnieri Assessment Office-of Nuclear Reactor Regulation

References:

(a) License No. DPR-3 (Docket No. 50-29)

(b) Letter, ljRC to Yankee Atomic Electric Company, dated August 31, 1990 (c) Letter, Yankee Atomic Electric Company to NRC, dated July 5, 1990

Subject:

Reactor Pressure Vessel Fluence Assessment Dear Sir As noted in the NRC Safety Assessment of the Yankee-reactor pressure vessel (Reference (b)), Yankee committed to' preparing and submitting an updated fluence analysis by Occober 1, 1990. -In a telecon with Dr. Thomas Murley, NRC, on September 25, 1990 Mr. John DeVincentis of Yankee Atomic Electric Company (Yankee) reported that preliminary results of the updated fluence analysis,-which is still under review, indicated a change from the fluence referenced in previous' analyses sent to the NRC. The preliminary results indicate a higher peak fluenes (i.e., 2.6 x 1019 versus 2.3 x 1019 2

n/cm ) and a higher azimuthal variation of fluence (i.e.,

1.2 x 1019 versus 3.7 x 1018 n/cm2 at 45').

Mr. DeVincentis reported that Yankee assessed the effect of the changes and determined that the resulting PTS analysis was still within the bounds of the NRC Safety Assessment'and that startup of the Yankee plant from its refueling outage was. therefore, justified.- Dr. Murley concurred with that judgment.

Mr. DeVincentis also informed Dr. Murley that Yankee would not be able to

meet the October 1 -1990 submittal date.for the final updated fluence analysis because further work had to be performed in order to verify the preliminary results and assess their effects.

In a subsequent telecon between NRC and Yankee on September 26, 1990, it was agreed that the safety assessment performed by_ Yankee on-the preliminary fluence values be submitted on September 28, 1990=and that the final updated fluence analysis would be submitted within 60 days of the telecon.

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United States Nuclear Regulatory Commission Septemoer 28, 1990 l

Attention Mr. William Russell Page 2 BYR 90-128 Safety Assessment A saf ety assessment was perf ormed to determine the ef f ects of the preliminary updated fluence values on previous PTS analyses.

The assessment first determined the 1990 reference temperatures f or the beltline materials.

Then, using these reference temperatures, PTS fracture mechanics analyses were performed f or each material to obtain their conditional f ailure probabilities.

The reference temperatures were determined using the updated fluence values.

The method for calculating the fluences in the axial and azimuthal directions was the same as described in Yankee's Reactor Vessel Evaluation Report (deference (c)).

The resulting fluence distribution is shown in Table 1.

The reference temperature values for the beltline materials were calculated using the methods contained in the NRC Safety Assessment.

1.

Upper Plate The reference temperature estimated by the NRC for the upper plate is 19 n/cm and the evaluation of based on a peak fluence of 2.3 x 10 C. R. Odette.

Odette e-tinated the shif t in reference temperature to be 245'F.

To be consistent with this approach, the trend curve for Yankee surveillance data has been increased by 60*F as shown in Figure 1 to reflect a 245'T shift at a 2.3 x 1019 n/cm fluence.

Application of the new fluence distribution from Table 1 and use of the revised trend curve.

Figure 1, results in the reference temperature distributien as shown in Table 2 for the upper plate.

2.

Lower Plate In the NRC Safety Assessment, the NRC estimated that the increased nickel content in the lower plate would contribute an additional 80*F above the revised trend curve, Figure 1, for the upper plate. Application of the new fluence distribution for the lower plate from Table 1. use of the revised trend curve, and addition of 80'T results in the reference temperature distribution for the lower olate as shown in Table 2.

3.

Circumferential and Axial Welds The NRC estimated the reference temperatures for the circumferential and axial welds using Regulatory Guide 1.99, Revision 2 methodology; bounding values for copper and nickel of 0.35 wt.% and 0.70 wt.%, respectively; and 50*F for the irradiation temperature effect. Application of the updated fluence distribution for the welds from Table 1, use of Regulatory Guide 1.99, Revision 2 methodology, and addition of 50*F results in the reference temperature distributions for the welds as shown in Table 2.

The reference temperatures for each material were input into the VISA II Code as before.

The distribution of input parameters, flaw density using the Marshall distribution, flaw length, assumption of one flaw per beltline material, and number of simulations were identical to previous submittals to t he NP.C.

i i

Unitod States Nuclear Regulatory Commission September 28, 1990 Attention:

Mr. William Russell Page 3 BYR 90-128 The f racture mechanics were run with the limiting thermal hydraulic parameters f rom all three accident categories (small break LOCA, main steam line break, and transient).

The small break LOCA was dominant because it had the highest combination of conditional failure probability and event frequency. As provided in Table 3, the results indicate e conditional failure probability of 2.75 x 10-2 The limiting material is the upper axial weld.

The results are 9till within the bounds assumed by the NRC in its Saf ety Assessment, i.e., 10-' to 10-2 for conditional failure probability. There are some conservatisms used in the calculation for the axial welJ which should be taken into account in assessing the results.

First, one flaw is assumed in each axial weld for fracture mechanics.

The volume of the weld material is very small, and one flaw in the upper axial weld is equivalent to approximately 50 flaws per cubic meter.

Second, the axial welds are located between the cold leg nozzles which see higher temperatures than below the nozzles. The beltline area, including the axial welds, was assumed to see the same cold temperatures for the analysis.

Third, the thermal hydraulle conditions assumed were stagnated flow; during the entire event which results in the coldest temperatures for the longest times.

Fourth, the axial weld reference temperatures are based upon bounding copper and nickel contents of 0.35 wt.% copper and 0.70 wt.1 nickel.

A sensitivity study was performed to assess the change in fracture mechanics results if the copper and nickel content were 0.30 wt.% and 0.70 wt.%, respectively.

The results show that the conditional f ailure probability decreases to 7.6 x 10-3 Therefore, the results are highly sensitive to the copper and nickel assumptions.

Based on the results described above, Yankee concludes that startup and operation of the plant are justified for the following reasonst 1.

It is very unlikely that Yankee would experience a PTS event. The design features on which this conclusion is based are not affected by the change in fluence.

2.

The PTS conditional failure probability with the new fluence distribution is within the bounds assumed by the NRC in its Safety Assessment.

3.

The upper axial weld is the limiting material for PTS.

Its reference temperature is based upon bounding copper and nickel content. The results are sensitive to the chemistry content, and if lower copper contents are identified, better results are obteined.

4 The conservatisms associated with the flaw density within the upper axial weld, the thermal hydraulic conditions assumed for the small break LOCA, and the location of the axial welds between the cold leg nozzles would provide margin for any remaining uncertainties.

Based on the above. Yankee concludes that reasonable assurance of the public health and safety continues to be provided.

United States Nuclear Regulatory Commission September 28, 1990 Attention:

Mr. William Russell Page 4 BYR 90-12 8 Additional Information As requested by the NRC, volume and inside surface areas of the axial welds, circ unferential weld, upper and lower plates are contained in Table 4.

,f4 Within 60 days we will submit final fluence values and will revise the PT.

analysis to account for the final values.

Sincerely, eb, 6

John D. Baseltine Director Yankee Project t

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Attachments cc:

P. Sears (NRC, NRR)

R. Wessman (NRC, NRR) t

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P Table 1 Fluence Distribution for BettLine haterlats Peak Fluence at 21.44 EFPT 2.6 e19 n/cm2 A2iMUTNAL VARIATf0A hist Welds o to 5 5 to 10 10 to 1515 to 20 20 to 25.5 to 30 30 to 35 35 to 40 40 to 45 40 to 45 Urte Plate 10 to 20 0.360 0.376 0.390 0.373 0.332 0.280 0.235 0.202 0.187 0.187 20 to 30 1.236 1.292 1.339 1.287 1.141 0.960 0.806 0.6 95 0.641 0.641 30 to 40 1.949 2.037 2.11'.

2.029 1.799 1.514 1.271 1.096 1.011 1.011 40 to 50 2.222 2.323 2.408 2.314 2.051 1.726 1.449 1.250 1.153 1.153

% sf Neight 50 to 60 2.335 2.441

~ 2.530 2.431 2.155 1.814 1.523 1.313 1.212 1.212 60 to 70 2.335 2.441 2.530 2.431 2.155 1.814 1.52' 1.313 1.212 1.212 70 to 80 2.400 2.509 2.600 2.499 2.215 1.864 1.565 1.349 1.245 1.245 80 to 90 2.357 2.464 2.553 2.454 2.175 1.831 1.537 1.325 1.223 1.223 90 to 100 2.335 2.441 2.530 2.431 2.155 1.814 1.523 1.311 1.212 1.212 Circ Cald 2.136 2.233 2.314 2.224 1.9 72 1.659 1.393 1.201 1.108 ISe^rFlate 35 to 40 0 to 10 2.136 2.233 2.314 2.224 1.972 1.659 1.393 1.201 1.108 1.201

% of Height 10 to 20 1.680

1. 756 1.820 1.749 1.551 1.305 1.096 0.945 0.872 0.945 20 to 30 0.?70 1.014 1.050 1.009 0.895 0.753 0.632 0.545 0.503 0.545 30 to 40 0.168, 0.176 0.142 0.175 0.155 0.130 0.110 0.094 0.087 0.094

e 9

1

• I Table 2 Mean Ref erence Tenperatures Based m New Fluence Calculation Peak Fluence at 21.44 EFPY =

2.6 e19 n/cm2 A21MUTHAL VARIATION Axle l Welcs 0 to 5 5 to 10 10 to 1515 to 20 20 to 25 25 to 30 30 to 35 35 to 40 40 to 45 40 to 45 Ugwr Plate 10 to 20 115 118 122 118 108 99

<96

<96

<96 175 20 to 30 203 205 208 205 197 185 172 163 157 248 30 to 40 235 137 240 237 228 210 205 195 188 277 40 to 50 243 245 250 245 236 225 213 203 197 285 g

% of Height 50 to 60 248 250 25 2 250 240 230 217 205 200 289 60 to 70 248 250 252 250 240 230 217 205 200 289 70 to 80 250 25t 255 252 243 231 218 2C8 204 290 80 to 90 247 250 25 3 250 241 230 217 206 202 289 90 to 100 245 248

?51 248 240 230 216 205 201 289 Circ Veld 323 326 328 326 319 308 297 288 283 35 to su Lower Plate u to 10 320 323 325 323 315 303 20.8 280 275 22

% of Height 10 to 20 304 306 310 305 298 285 275 265 25 8 2 73 20 to 30 265 257 272 267 260 248 23 5 225 220 238 30 to 40

<176

<176

<176

<176

<176

<176

<176

<176

<176 142 Notet To cetermine the ref erence temeratum, an initial temperature of 307 for plates and 10F for welds must be added to the r mean reference temeratures.

Table 3 PTS Fracture Mechanics Results 1990 Peak Fluence Peak Reference Conditional Failure

( x 1E+19 n/cm-2)

Temperature -F Probability Upper Plate 2.6 285 2.8 E-03 Lower Plate 2.31 355 5.6 E-05 Circ. Weld-2.31 338 6.8 E-04 Upper Axial Weld 1.24 300 2.4 E-02 Lower Axial Weld 1.2 298 2 E-05 2.75 E-02

+

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Table 4 Volume and Inside Surface Area of Beltline Materials Volume (ft.-3)

Inside Surface Area (in.-2)

Upper Plate 124 25434 Lower Plate 46 9470 Circ. Weld 3

300 i

Upper Axial. Weld 0.63 66 Lower Axial Weld 0.24 25 3

--,,7

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r

3 7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION YANKEE ATOMIC ELECTRIC COMPANY YANKEE (ROWE) NUCLEAR POWER STATION DOCKET NO.50-029 NOTICE OF-RECEIPT OF PETITION UNDER 10 CFR 2.206 Notice is,hereby given that the Director, Office of Nuclear Reactor Regulation,. acknowledges receipt of a petition filed jointly by the Union of Concerned Scientists and the New England Coalition on Nuclear Pollution pursuant to 10 CFR 2.206 for emergency enforcement action against Yankee Atomic Electric-Company's Yankee <Rowe Nuclear Power Plant.

The petition seeks'the immediate shutdown of the Yankee-Rowe Nuclear Power Plant which the petitioners allege are ~ operating in violation of the Nuclear Regulatory Commission's standards for pressure vessel integrity.

For the reasons

- discussed in a letter to Diane Curran from Thomas E. Murley, dated June 25, 1991, the request-for immediate relief has been denied.

A decision ~ concerning this petition will be ' addressed in a final dechion

.in the near future.

~

A copy of the petition is available for public inspection in the Commission's PublicIDocumentRoom,locatedintheGelmanBuilding,2120LStreet,N.W.,

Washin'gton, D.C.

20555.

Dated at Rockville, Maryland,'this 25th day of June,1991.

FOR THE NUCLEAR REGULATORY COMMISSION Thomas E. Murley, Director Office of Nuclear Reactor Regulation 1.

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ORIGI!JAL UUE UT: v 7.1,

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PHHTLOW L RF E V AF:t:r.

HUSSELL blLLESFIE I arJ4 e

F J. e ra s e r- (9 v i ( > ui t.h e (j ui-cl a t. e 1erone at at.ei :

1i t:he due d es t e d oo s: not 911aw adequate

t. I me ta respond ca thts t t C + Q 1", yOU F) a y r" (P C}U e s t.

3 r F.a v l B e d t'I t t E4 d a t. e.

The -maaest must have prior opprous1 ' r o.n th+

c p pr opt tate A,r.uctatf 01rei: tar or NRR Deputy Di re : ti > ~ and must t nc l i to,e a v :i l 2 d Justiftcetion.

Cantact NRR ins t t 1. n:w w i t I, i *,e n i.w di s > ci cA t,

• it;e1 est e Sili v r e, em t --212.f9 ).

l> 10a sG: (10 nat car-contutrtsnte nr.u,bgem to U t r ec lar c3 office W 1 t l)Di t t.

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n ACTION DUE TO NRR DIRECTOR'S OFFICE BY __f /5' &

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