ML20052D144

Forwards Info Re Pressurized Thermal Shock,Including Discussion of Cooldown Events Experienced at Seven Operating Plants w/C-E Rcs,In Response to NRC 820318 Request for Addl Info

ML20052D144
Person / Time
Site:	Calvert Cliffs
Issue date:	05/04/1982
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
REF-GTECI-A-49, REF-GTECI-RV, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-A-49-.13, TASK-OR, TASK-TM NUDOCS 8205060310
Download: ML20052D144 (13)

Text

,

z

. BALTI MORE \ 'q, GAS AND ELECTRIC U([\ C-h ti: ..

-l Q'

CHARLES CENTER P. O. BOX 1476. BALTIMORE, MARY D 21203 D ' d, ?s a j

'\ q , c3 >

ARTHUR E. LUNDVALL. JR.

M '4,l1982

% q,,y[#

vice pars ocm sumv Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Mr. D. G. Eisenhut, Director Division of Licensing

Subject:

Calvert Cliffs Nuclear Power Plant Unit No.1, Docket No. 50-317 Pressurized Thermal Shod < (PTS)

Reference:

(a) Letter from R. A. Clark to A. E. Lundvall, Jr. dated March 18, 1982 (b) Letter from A. E. Lundvall, Jr. to D. G. Eisenhut dated Janua.y 28, 1982 Gentlemen:

Reference (a) requested the submittal of additional information on the subject of PTS be provided by April 30,1982. This letter provides our response.

Enclosure (1) provides discussions of each of the topics on which additional information was requested. Enclosure (2) is a discussion of cooldown events experienced at seven different operating plants with C-E reactor coolant systems.

We trust you will find this information responsive to your request. Based on the evaluations previously submitted and considering the additional information provided here, there is no near-term problem with pressurized thermal shock at Calvert Cliffs Nuclear Power Plant Unit 1. We stress that an orderly program for resolution of this concern is appropriate.

Very truly yours, l -

R. C. Brya ~E all, Jr.

fora.E.Lund[

Vice Presi dent-Supply cc: J. A. Biddison, Jr., Esq.

G. F. Trowbridge, Esq.

D. H. Jaffe - NRC y R. E. Architzel - NRC 8205060310 820504 PDR ADOCK 05000317

" i1 P PDR

~

o

• ,Enclos*ure 1 The first two questions addressing operator action concerned CEN-189, which was submitted in response to Action Item II.K.2.13 of NUREG-0737. This item required f that a detailed analysis be performed of the thermal-mechanical conditions in the l

reactor vessel during recovery from small breaks with an extended loss of all feedwater. The recommended mode of decay heat removal for the CE NSSS is by means Therefore, loss of all feedwater as re-of the steam generator secondary side system. Since the i

I quired by II.K.2.13 represents a potential inadequate core cooling situation.

concern in ll.K.2.13 is the therma!-mechanical conditions of the vessel, only scenarios which permit adequate core cooling were considered, in order that " recovery" could be assured. Accordingly, CE reviewed CEN-ll4, "Small Break Transients in CE NSSS's,"

which was submitted to the NRC in 1979, and chose two different basic modes of

" recovery" from loss of all feedwater.

One mode of recovery selected from CEN-ll4 for PTS evaluation was to re-establish auxiliary feedwater. The CE U-tube steam generators have sufficient secondary side inventory such that 15 to 30 minutes are required to dry out a steam gen-Reestablishing feedwater prior to dryout would erator af ter loss of all feedwater.

represent a less severe PTS situation due to mixing of the auxiliary feedwater with the The 30 minutes chosen for re-establishing remaining contents of the steam generators.

feedwater, as reported in CEN-IS9, represents the time af ter which the steam gen-erators would be essentially dry and therefore this transient was basically an overfeed event to dry steam generators.

The other mode of recovery selected for PTS evaluation was to open PORV's to prevent inadequate core cooling, assuming a total and continued loss of all feedwater.

This method of cooldown is not advocated by CE but was evaluated simply to satisfy the requirement of II.K.2.13 to assume loss of all feedwater. The operator action of opening the PORV's at 10 minutes is completely dictated by the core cooling aspects of this scenario, and is not subject to variation for PTS consideration.

Since the parameters for the two types of cases reported in CEN-189 were chosen to maximize the PTS aspects of the transients (subject to the limits of maintain-ing adequate core cooling), and since these cases were found to be less severe than the cases reported in reference (b), we believe sufficient variation of the CEN-189 cases has been accomplished.

The third question addressing operator action concerned the sensitivity of the MSLB analysis to the time assumed for operator action. Reviewing pressure-temperature )

transients with RCP trip times of 30 seconds and 5 minutes, it was found that the cooldown transient is less severe for the later RCP trip case. Fracture mechanics evaluations performed using forced convection heat transfer coefficients during the pumps-on portion of the cooldown transients confirm that the early pump trip case is more limiting than the delayed pump trip case. Therefore, the assumed pump trip time reported in reference (b) is conservative.

Additional evaluations were performed to evaluate sensitivity to the time assumed for the operator to restrict ECCS flow. It was found that the conditions for minimum subcooling to assure no core voiding are achieved early in an MSLB transient,

F O

such that prompt action would effectively prevent any repressurization. The MSLB t ansients reported to date show repressurization would begin about 6 to 8 minutes after start of the event. Repressurization to the HPSI pump head is observed about 10 to 12 minutes into the event. If action is taken to prevent repressurization to the HPSI pump head, then many additional EFPY of acceptab'e performance can be demonstrated.

Action later in the transient would still have the benefit of ensuring depressurization.

We have evaluatsd the need and effectiveness of procedure modifications to clearly identify reactor vessel integrity concerns. We agree that it is crucial that this aspect of the PTS concern be properly addressed. Since our letter of October 20, 1981, we have implemented changes to our procedures which serve to remind the operator of presure and temperature limits to be observed in the interest of . reactor vessel integrity. These are provided at points in the procedure where the operator would be taking action to ensure adequate core cooling. Information on PTS has been disseminated to our operators. Formal training of all operators will be completed by the end of June 1982. This training complements the changes to procedures which have already been m ade. We are continuing our efforts under the CEOG to develop improved emergency procedures guidelines which properly integrate PTS concerns. We expect to have implemented these guidelines by October of this year.

Concerning your request for plant experience with overcooling events, a generic CEOG task was performed to identify events which have occurred at operating plants with a C-E NSSS. Events which resulted in a cooldown rate in excess of 100 F/hr, re-suited in a cooldown of at least 1000F, and had a duration of more than 10 minutes were reviewed. At Calvert Cliffs, preliminary review of our entire operating history (for both units) resulted in identification of more than 70 events for further evaluation. Screening these events yielded a list of three events app opriate to the CEOG effort. Sixteen candidate events were identified by a review of seven operating plants. Six of the events selected satisfied none of the criteria; three events satisfied only the 100 F/hr criterion; one event satisfied the 100 F/hr criterion for mo e than 10 minutes, and only two events satisfied all three criteria. (Detailed information on the remaining four events was not available in time for this report.)

Of the three events which satisfied at least two criteria, the longest duration of rapid cooldown was 19 minutes and the maximum Tave temperature decrease was 107 F. None of the three events exhibited repressurization at low temperature.

Enclosure (2) provides the specific results of this review.

In response to your request for probabilistic risk assessment of potential over-cooling events, a review of available data was all that could be accomplished in the time available. An extensive list of possible PTS scenarios was considered including different types of initiating events at different plant operating conditions. Specific event-plant condition combinations judged to have a high likelihood to lead to the most severe PTS sequences were chosen for detailed study. The sequence of events for each of the selected combinations was determined using the sequence tables and diagrams in FSARs and CEN-128. Probabilities were determined for all logical, relevant scenarios. The l scenarios which resulted from this effort were categorized as Moderate Frequency (50%

l probability of occurring in any one year); Infrequent (50% probability of occurring once during plant 40-year lif teime); and Limiting Fault-1, -2, or -3 (low, very low, or exceedingly low probability of occurring during plant 40 year lifetime).

l 2 1

(

L  :.

In summary, no Moderate Frequency or Infrequent events were determined among the scenarios considered. The MSLB initiating event was categorized as a Limiting Fault-3 event. There were a few Limiting Fault-1 and -2 events identified, but none was judged to represent a more challenging PTS event than the lower probability MSLB event. .

Although not specifically requested, we have new . chemical composition information obtained through EPRI concerning the critical wends in Calvert Cliffs Nuclear Power Plant, Unit No.1. This information indicates that the copper and ni&el contents assumed in all submittals to date has been overly conservative. This new chemical composition information is based upon the . chemical analysis of other weldments produced with the same welding wires and fluxes as those used in the Calvert Cliffs vessel.

The new information is currently under review by BG&E and a report concerning the impact of this information will be issued soon. A preliminary assessment of this new data shows that the welds wili embrittle at a rate much lower than previously assumed.

As a result, the design life of Calvert Cliffs Nuclear Power Plant, Unit No. I should not be reduced as a result of pressurized thermal shod consideratinr.:.

3 L

Enclosure 2 Pressurized Thermal Shock Precursor Events of Operating Nuclear Plants -

. with a C-E NSSS 1.0 PURPOSE The purpose of this report is to docunent Pressurized Thermal Shock (PTS) precursor events that have occurred at nuclear power plants with a C-E supplied NSSS. This report is intended to provice additional scoping information on PTS for the C-E Owners Group.

2.0 SCOPE The scope of this report is limited to the identification of PTS precursor events that have occurred at the following seven operating nuclear power plants with a C-E supplied NSSS: Palisades, Calvert Cliffs 1 and 2 Ft.

Calhoun, Arkansas Nuclear One Unit 2, Millstone 2, and St. Lucie 1. This report covers C-E operating experience from January,1971 through February, 1982.

3.0 REFERENCES

C-E Reliability Data System.

1)

2) "LER Monthly Report Sorted by Facility for Power Reactors" U.S.

Nuclear Regulatory Commission, Washington, D.C.

3) NUREG-0020, " Operating Units Status Report - Licensed Operating Reactors - Data for Decisions" U. S. Nuclear Regulatory Commission, Washington, D.C.

4) " Nuclear Power Experience", Nuclear Power Experiences, Inc.

i i ,

5) Herbst, J. J., Paggen, V.A., " Combustion Engineering Availability l Program for Nuclear Steau; Supply Systems", presented at 33rd Annual ,

Technical Conference of the American Society for Quality Control Nuclear Division, May 14 - 16, 1979, Houston, Texas.

6) MN-82-08, J. B. Randazza (Maire Yankee) letter to R. A. Clark (NRC),

dated January 21, 1982.

7) A. E. Lundvall (BG&E) letter to D. G. Eisenhut (NRC), dated January 28, 1982.

8) LIC-82-029, W. C. Jones (0 PPD) letter to T. M. Novak (NRC), dated January 18, 1982.

4.0 BACKGROUND

In March,1982 the NRC requested that the C-E Owners Group provide additional material to support the CE0G position on PTS. One specific item requested by the NRC was a list of PTS precursor events that have occurred at nuclear plants with a C-E supplied NSSS. This report provides that requested information.

5.0 METHODOLOGY Identification of actual PTS and PTS precursor events at operating nuclear power plants involves determining the conditions that existed in a plant during each transient at that plant and comparing these conditions to a predetermined set of criteria for selecting PTS precursor events.

p- - _ _ _ _ - - . . . _ _ =

The identification process is based on four steps including:

o Development of PTS precursor criteria o Research and preliminary screening of events o Confirmation of event details o Identification of PTS precursor events This process is further described below.

5.1 Step 1 - PTS Precursor Criteria Development The first step in the process was to select criteria for judging whether or not a specific event was a Pressurized Thermal Shock precursor. The basic PTS precursor criteria are:

(1) > 100 F/ hour cooldown rate, and (2) > 100 F total cooldown, and (3) > 10 mins. duration s to allow for reactor vessel response.

Application of these criteria requires a certain amount of detailed information about each transient, information that, in general, is only available at the plants. The following criteria were developed for preliminary screening of transients:

(a) the transient involved excess steam flow, or (b) the transient involved excess feedwater flow, or (c) the transient involved a decrease in RCS pressure followed by actual safety injection flow.

5.2 Step 2 - Preliminary Screening Available operating experience information sources (1, 2, 3, 4, 5) were reviewed to identify events at C-E plants which met the preliminary  !

screening criteria for PTS precursor events. A brief description was written for each such event. This preliminary review covered a total of approximately 49 plant years of operating experience. f l _ _ _ = -

(

5.3 Step 3 - Acquire Detailed Event Data Detailed information was requested from the appropriate utility for each event selected in the preliminary screening. Confirmation of potential PTS  ;

precursor events was requested directly from the plants by C-E personnel .

5.4 Step 4 - Identify PTS Precursor Events Using the detailed information acquired in Step 3, each potential PTS ,

precursor event was evaluated against the basic PTS precursor criteria to see if it was a PTS precursor. A summary of events which met the basic PTS precursor criteria are given in Table 2.

6.0 RESULTS r

Sixteen potential precursor PTS events were identified in the screening of operating experience data for'seven operating plants with a C-E supplied NSSS. Table 1 contains a brief description of each of these events.

Detailed information was provided for the potential PTS precursor events by the utilities. Evaluation of the detailed event information against the i -

basic PTS precursor criteria produced the following results:

(a) 6 events met none of the three criteria, (b) 3 events met one criterion, temperature drop > 100 F/ hour, (c) 1 event met two of the criteria, temperature drop >100 F/ hour and duration > 10 minutes, (d) 2 events met all three criteria, (e) 4 events do not contain sufficient information to evaluate. l Table 2 provides additional data for the two events that met the PTS  !

I precursor criteria.

_h_

The concluslon of this study is that in 49 plant years of operating experience, nuclear plants with C-E supplied NSSSs have experienced only twp events that met the basic PTS precursor selection criteria. Both events were significantly less severe, in terms of temperature drop, than the cases analyzed for the 150 day submittals (7, 8, 9). In neither case did the plant repressurize as a result'of the event.

e t

O

=

G e

1 l

' - I,

, g Table 1 .

POTENTIAL PTS EVENTS FROM PRELIMINARY SCREENING PLANT DATE ,

DESCRIPTION Il

)

Palisades 12/16/78 Reactor tripped on low steam generator l (SG) water level . Main feed pump "B" ,

failed to trip causing overfeed ,{

transient. RCS cooldown caused a safety l ,

injection (SI). Feed pump subsequently tripped by operator. ,

Palisades 02/01/79 Inadvertent trip of ar. RCP by operator caused a reactor trip. Overfeed event, apparently caused by a .' ailed open main feedwater regulating valve, caused an RCS cooldown. Safety Injection resul ted. Feedwater regulating valve manually closed and feed pump tripped to terminate transient.

Calvert Cliffs-1 05/10/75 During full power testing two turbine '

bypass valves stuck open. Resultant RCS cooldown caused a $1. RCS temperature -

reportedly decreased approximately

. 100*F over several minutes.

Approximately 2500 gallons injected

, over 10 minutes. Transient terminated when operators were dispatched and manually closed the stuck open turbine bypass valves.

Ft. Calhoun 04/74 An inadvertent loss of main feedwater -

caused a reactor trip on low SG level. j A turbine bypass valve was subsequently opened to facilitate heat removal .

Overfeeding of one steam generator occurred due to a stuck open feedwater regulating valve. RCS cooldown caused a SI. Natural circulation cooldown initiated when SIAS isolated CCW to RCP seals. Transient terminated by manually closing the affected valves.

ANO-2 12/27/78 A main steam relief valve lifted and failed to reseat during turbine roll at  !

near hot zero power (HZP) caused a 107 F l

RCS cooldown over 52 minutes. Relief valve reseated after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> blowdown lowered pressure.

I l

Table 1 (Continued)

POTENTIAL PTS EVENTS FROM PRELIMINARY SCREENING e PLANT DATE DESCRIPTION ANO-2 01/79 Similar to ANO-212/27/78 event. Main steam relief lifted and failed to '

reseat. Reactor manually tripped but blowdown continued for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Valve subsequently reseated without operator action.

Ft. Calhoun 12/77 During RCS cooldown, a SI occurred '

because SIAS was not blocked prior to reaching actuation setpoint. SIAS subsequently reset.

Palisades Precommercial During precommercial testing, two RPS ,

breakers were opened simultaneously causing PORVs to open. Safety Injection resul ted. The control room operator shut the PORV block valves to terminate the transient.

Mill stone 03/80* Loss of a main feedwater pump caused a  !

reactor trip on low SG water level. The "B" condenser dump valve stuck open on the trip. Steam dump control was placed in manual and the dump valve was closed ,

to terminate the transient.

Calvert Cliffs-2 12/76 During operator training at 19% power, two SG overfeeding events occurred, each ,

apparently resulting in an RCS cooldown. Transient terminated by manual reactor trip and manual control of the feed regulating valves.

Mill stone-2 02/26/76 A dropped rod occurred from 80% steady state power. Erratic control of SG water levels in manual apparently caused i a reactor trip. Main feedwater ramped  ;

back to 50% of normal full flow, turbine .

runback apparently did not occur and the "A" steam dump valve stuck open momentarily. The feedwater regulating valves were shut in manual control and l the steam dump reseated without operator intervention to terminate the transient.

l l

1

-T-

Table 1 (continued)

POTENTIAL PTS EVENTS FROM PRELIMINARY SCREENING PLANT DATE DESCRIPTION Calvert Cliffs-2 04/12/81 While returning to full power after condenser repairs, operator inadvertently overborated RCS. Reactor

. power sharply decreased. Reactor reportedly tripped due to insthbilities in SG water level control.

Calvert Cliffs-2 09/21/81 Reactor manually tripped at full power  ;

in response to a break in a main {

feedwater line. Feedwater transient not j seen by steam generators. l l Mill stone-2 01/02/81 The reactor tripped from full power due I

to the loss of one 125V DC bus. Turbine trip was delayed approximately 30 seconds, resulting in an RCS cooldown.

Turbine manually tripped at local control board to terminate transient. -l ANO-2 01/29/80 Turbine was tripped from 100% load for

. power ascension tests. Reactor tripped on low SG level due to shrink effect.

One steam dump valve stuck open causing I

a cooldown. A pressurizer spray valve

, also stuck open causing RCS pressure to decrease and SIAS actuation. The steam dump was manually shut and a containment I _ entry was made to " gag" shut the spray

> val ve.

Palisades 02/04/82 Reactor tripped on thermal margin / low pressure during a rapid power de-escalation following loss of "A" cooling tower pump. Spray valves opened, i

secondary safeties opened and steam dump valves opened. Secondary safeties and steam dump valves subsequently closed after extensive blowdown without operator intervention.

• Insufficient information to evaluate.

t ..

TABLE 2 .

~

EVENTS AT PLANTS WITH C-E NSSSs

{ MEETING PTS PRECURSOR CRITERIA -

Number }

Maximum Rate Duration of Total RCS Repressurize of i

Plant Date of Temperature Maximum Temperature Temperature Pressure at low Criterion D crea e Change Rate reas Decrease Temperature Met Tave -DeIave)e

(

U Ft. Calhoun 4/74 .330 F/hr. 19 Minutes 107 F 700 psig No 3

Arkansas 2 12/78 156 F/hr. 10 Minutes 107 F ~900 psig No 3 i

e I

I l

! Criterion 1: Criterion 3: Criterion 2:

! >100 F/hr. >10 minutes >100 F cooldown cooldown rate event duration - of RCS 1 .

1

_ _ _ _ . - _ __.