ML20137F240
| ML20137F240 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 03/25/1997 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M97855, TAC-M97856, NUDOCS 9704010019 | |
| Download: ML20137F240 (17) | |
Text
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, c CHARIIS II. CRUSE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4455 March 25,1997 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Second Request for Additional Information: License Amendment Request;
/ Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging (TAC Nos. M97855 and M97856)
REFERENCES:
(a) Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated January 31, 1997, License Amendment Request; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging (b) Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE),
dated March 5,1997, Request for Additional Information - Proposcd Technical Specification Changes to Reactor Coolant System Flow Limit, Calvett Cliffs Nuclear Power Plant, Units I and 2 (TAC Nos. M97855 and M97856)
By Reference (a), Baltimore Gas an6 Electric Company submitted a license amendment request to the Nuclear Regulatory Commission (NRC) to support operation of Calvert Cliffs Units 1 and 2 with up to 2500 steam generator tubes plugged in each steam generator. By Reference (b), the NRC requested additbnal information regarding the license amendment request. Additional NRC questions were presented by telephone on March 12, 1997. Attachment (1) provides Baltimore Gas and Electric Company's response to the questions posed in Reference (b), and Attachment (2) provides the response to those presented by telephone on March 12.
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Docum[nt Control Desk March 25,1997 l - Page 2 l This additional information does not change the Significant Hazards Determination presented in l Reference (a). Should you have further questions regarding this matter, we will be pleased to discuss them with you.
l Very truly yours,
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STATE OF MARYLAND :
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- . TO WIT:
COUNTY OF CALVERT :
I, Charles H. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this
, License Amendment Request on behalf of BGE. To the best of my knowledge and belief, the statements
! contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other BGE employees and/or
- consultants. Such information has been reviewed in accordance with company prac 'ce and I believe it to i
be reliable.
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Supsc ibed and sworn before me, a Notary Public in and for the State of Maryland and County of L , this X5tkiday of 7hdALN) .1997.
l WITNESS my Hand and Notarial Seal: b M(t Notary ) b PublicMAf.AA l.
My Commission Expires: IIMAAM Mi I9i (ate CllC/NH/ dim l
Attachments (1) Response to Request for Additional Information
! (2) Response to Additional NRC Questicas (As Provided by March 12,1997 Telephone Conference)
! cc: D. A. Brune, Esquire H. J. Miller, NRC J. E. Silberg, Esquire Resident inspector, NRC
. - . . -- -. .. . . .. . . - . . . - . . . . . . . - - . - . . . ~ - . . . _ . - . . . . - . - .
ATTACHMENT (1)
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i RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION l
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1 Calvert Cliffs Nuclear Power Plant Units 1 & 2 March 25,1997 i
, AITACIIMENT (1)
RESPONSE TO REQUEST FOR ADDFIIONAL INFORMATION NRC Question No.1 All revised accident analyses must assess the consequences of the accident with respect to General Design Criteria (GDC) 19. Please transmit the assumptions and dose associated with your analyses to demonstrate that GDC 19 is met.
BGE Response Baltimore Gas and Elec tric Company's (BGE's) interim analysis to demonstrate our current methods for complying .vith GDC 19, pending the completion of our thyroid dose calculation, was submitted by Reference (1). The interim habitability analysis indicates that our post-Loss-of-Coolant Accident thyroid dose will be maintained below the (GDC 19) 30-day limit of 30 rem, if the planned protective measures are implemented within a reasonable time frame (i.e.,45 minutes). The NRC approved this interim Control Room habitability analysis by Reference (2).
Two Design Basis Accidents, Main Steam Line Break (MSLB) and Seized Rotor Event, were reanalyzed in support of Reference (3). A preliminary analysis of Control Room habitability for these Design Basis Accidents has been performed using the AXIDENT Code, Reference (4),
which has been used by BGE for other dose calculations. AXIDENT was approved by the NRC in Reference (5). The key assumptions used in this analysis are presented in Table (1). We plan to finalize the Control Room habitability analysis for these events concurrent with completion of the final Control Room habitability dose analysis.
For the MSLB Event, using the current analytical methods and assumptions for equipment and performance, the evaluation showed that BGE would meet the GDC 19 limits for operator thyroid dose by implementing personnel protective measures, as per Reference (1), within one <
I hour of the initiation of the event. These results are acceptable, as they are bounded by those for the Loss-of-Coolant Accident thyroid dose analysis submitted by Reference (1). The whole body dose for the MSLB Event was determined to be less than 1.45 rem, which is within the acceptance criteria for GDC 19. For the Seized Rotor Event, the evaluation showed that, using the current analytical methods and assumptions for equipment and performance, BGE meets the GDC 19 limits with a 30-day thyroid dose of 17.0 rem, and a 30-day whole body dose of less than 0.30 rem. j NRC Ouestion No. 2 The January 31,1997 submittal contained no dose assessments for either the Steam Generator Tube Rupture (SGTR) or the rod ejection accidents, yet it would appear that the consequences would be increasedfrom previous accidents as a result of the increasedplugging. Provide the assumptions and the dose assessmentsfor the SGTR and the rod ejection accidents with the increased steam generator tubeplugging.
BGE Resnonse As discussed in Reference (3), the changes associated with steam generator tube plugging do not have a significant effect on the SGTR Event or the Control Element Assembly (CEA) Ejection Event. Qualitative evaluations were performed which considered the effects of tube plugging on 1
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- ATTACHMENT m I 1
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION these events. In both cases, the evaluations concluded that the appropriate NRC acceptance criteria for each event were met. Therefore, the analyses of record (including dose assessment) I for these events were not revised.
For the SGTR Event, an evaluation was performed to estimate the effect of steam generator tube plugging on the ruptured tube leakage rate, the amount of Reactor Coolant System (RCS) leakage that flashes to steam, and the mass of steam released. The evaluation concluded that the site boundary dose would increase slightly, but would remain less than 10% of the 10 CFR Part 100 guidelines. Ten percent of the 10 CFR Part 100 guidelines is the NRC's acceptance criteria for this event for Calvert Cliffs, as documented in Reference (6).
The current analysis of record for Calven Cliffs' SGTR is documented in Reference (7). The analysis assumes operator actions taken in conservative time frames, as determined by Emergency Operating Procedures and validated by plant simulator experience. The two-hour site boundary dose is based on an integrated calculation of steam releases and steam activity.
The significant assumptions used in the analysis of record are provided in response to Question 2 in Attachment (2).
For the CEA Ejection Event, the Hot Zero Power and Hot Full Power cases were evaluated to assess the impact of reduced RCS flow on the fuel performance. References (8) and (9) contain the results of the Hot Zero Power and Hot Full Power CEA Ejection Event analyses of record, respectively, for Calvert Cliffs. These analyses conclude that no clad damage will occur. The CEA Ejection Event evaluation for steam generator tube plugging addressed the proposed change in RCS flow on the previously calculated fuel performance results, and confirmed that the peak rod average enthalpy remained below the NRC acceptance criterion of 280 cal /gm, and that the previously reported clad damage results remain valid.
Although a detailed radiological dose assessment of a postulated CEA ejection was not performed for the proposed reduction in RCS flow, the previously reported doses were reviewed for continued applicability to reduced flow conditions. Reference (10) contains radiological dose assessment information from the last CEA Ejection Event analysis (Calvert Cliffs Unit 2, Cycle 8) that required NRC reuew and approval. That dose calculation was performed with an assumed value of 10% fuel failure at the NRC's request. The key assumptions employed in that radiological assessment of the CEA Ejection Event are unaffected by the reduction in RCS flow.
Therefore, the previously reported radiological consequences are valid and remain well within 10 CFR Part 100 criteria.
REFERENCES:
(1) Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated May 6,1993, Control Room Habitability - Interim Engineering Analysis for Thyroid Dose (2) Letter from Mr. D. G. Mcdonald, Jr. (NRC) to Mr. R. E. Denton PGE),
dated June 22,1995, Control Room Habitability Interim Analysis for 'Ihyroid Dose, Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2 (TAC Nos.
M86396, M86397, M36437, and M86438) 2
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (3) Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated January 31,1997, License Amendment Request; Change to Reactor Coolant System Flow Requirements to Allow Increased Steam Generator Tube Plugging (4) S. J. Nathan, "AXIDENT: A Digital Computer Dose Calculation Model,"
NUS-1954, Revision 3, February 1984 (5) Letter from Mr. D. H. Jaffe (NRC) to Mr. J. A. Tiernan (BGE), dated February 20,1986, Containment Vent System (6) Letter from Mr. D. H. Jaffe (NRC) to Mr. A. E. Lundvall, Jr. (BGE), dated June 24,1982, Amendment No. 71 to Calvert Cliffs Unit 1 Facility Operating License No. DPR-53 (7) Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis Report, Revision 20 (8) Letter from Mr. A. E. Lundvall (BGE) to Mr. R. A. Clark (NRC), dated l October 15, 1982, Amendment to Operating License DPR-69, Fiflh Cycle License Application (9) Letter from Mr. J. A. Tiernan (BGE) to NRC Document Control Desk, dated February 6, 1987, Request for Amendment, Eighth Cycle License Application (10) Letter from Mr. J. A. Tiernan (BGE) to NRC Document Control Desk, dated March 27,1987, Unit 2 Cycle 8 Reload - Request for Additional Information l
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, ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TABLE (1) - ASSUMPTIONS USED FOR GDC 19 EVALUATION Assumption Maximum Main Steam Line Seized Rotor Event Description Hypothetical Accident Break Failed fuel fraction 100 % 10 % 5%
Pretrip power level 102 % 102 % 102 %
Power peaking factor 1.0 1.7 1.7 Dose conversion ICRP-2 ICRP-2 ICRP-2 factors Source terms TID-14844 TID-14844 TID-14844 Fractional release 50% of alliodines and 10% for all isotopes 10% for all isotopes 100% of all noble except 12% for I-131 except 12% for I-131 gases in the core and 30% for Kr-85 and 30% for Kr-85 Secondary I-131 N/A 1.E-7 Cilgm 1.E-7 Ci/gm activity Primary 1-131 activity N/A 1.E-6 Ci/gm 1.E-6 Ci/gm Primary noble gas N/A 1.E-4/Ebar Ci/gm 1.E-4/Ebar Ci/gm activity Control Room volume 166,000 ft' 166,000 ft' 166,000 ft' RCS mass N/A 384,260 lbm 38 !,260 lbm Steam generator mass N/A 125,250 lbm per steam 125,250 lbm per steam generator generator Primary to secondary N/A 100 gpd per steam 100 gpd per steam leak rate generator at 70*F for generator at 70*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 19,944 seconds 3 3 3 Site boundary 1.30E-4 sec/m 1.30E-4 sec/m 1.30E-4 sec/m dispersion coefficient Release pathways: 0.20% containment 100% through main 100% through gg .3,em V lume Per day for steam gooseneck for Atmospheric Dump first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0-2 min Valves (ADVs) 0.10% containment 50% through main I 100% through ADVs V lume Per day steant gooseneck and Long Term thereafter 50% through ADV thereafter ,
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, ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION Assumption Maximum Main Steam Line Seized Rotor Event Description 11ypothetical Accident Break 3 3 3 Control Room 7.7E-4 sec/m 9.66E-4 sec/m 1.14E-4 sec/m dispersion - for 0 hr - 8 hr - for 0-2 min - for 0 hr - 8 hr coefficients 3 3 4.5E-4 sec/m 5.40E-4 sec/m' 6.65E-5 sec/m
- for 8 hr - 24 hr - for 2 min-8 hr - for 8 hr - 24 hr 3
2.5E-4 sec/m' 3.15E-4 sec/m' 4.23E-5 sec/m
- for 24 hr - 96 hr - for 8 hr-24 hr - for 24 hr - 96 hr 3
6.3E-5 sec/m 2.00E-4 sec/m' l.82E-5 sec/m'
- for 96 hr-720 hr - for 24 hr-96 hr - for 96 hr-720 hr 3
8.61E-5 sec/m
- for 96 hr-720 hr Control Room inflow 21,100 cfm 21,100 cfm 21,100 cfm rates - for 0 sec < t < 2.1 sec - for 0 sec < t < 2.1 see - for 0 sec < t < 2.1 sec 11,005 cfm 11,005 cfm 11,005 cfm
- for 2.1 sec < t < 25.4 - for 2.1 sec < t < 25.4 - for 2.1 sec < t < 25.4 sec sec sec 910 cfm for t > 25.4 sec 910 cfm for t > 25.4 sec 910 cfm for t > 25.4 see Time to release N/A All secondary activity All secondary activity secondary activity released in 40.1 see for released choked flow at 830 instantaneously at t = 0.
psia.
Control Room A single filter initiates A single filter initiates A single filter initiates filtration at 2.1 see assuming at 2.1 see assuming at 2.1 see assuming 2,000 cfm at 90% 2,000 cfm at 90% 2,000 cfm at 90%
efficiency, efficiency, efficiency.
A single failure negates A single failure negates Second filter initiates at the second filter train. the second filter train. 1200 sec. assuming 2,000 cfm at 90% eff.
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1 i l l ATTACHMENT (2) i i
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RESPONSE TO ADDITIONAL NRC QUESTIONS i
I (As PROVIDED BY MARCH 1.2 , 1 9 9 7 TELEPHONE CONFERENCE) l 1 l l
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Calvert Cliffs Nuclear Power Plant Units 1 & 2 March 25,1997
, ATTACHMENT (2)
, , RESPONSE TO ADDITIONAL NRC QUESTIONS (As Provided by March 12, 1997 Telephone Conference)
Our response to the questions presented by telephone on March 12,1997 are provided below.
(Note: For the Steam Generator Tube Rupture and Control Element Assembly Ejection Events, the questions below are answered based on the analyses of record for these events, References (1) and (2),
respectively. These analyses were not revised to support the effort to allow plugging of up to 2500 tubes in each steam generator.)
- 1. For the Main Steam Line Break event, provide:
(a) The mass ofliquid releasedfrom thefaulted steam generator.
RESPONSE 125,250 lbm of steam is released from each steam generator. The entire initial activity from each steam generator is assumed to be instantaneously released to the environment for the site boundary dose calculation.
(b) The mass ofsteam releasedfrom the intact steam generator as afunction oftime.
RESPOSSE: 125,250 lbm of steam is released from each steam generator. The entire initial activity from each steam generator is assumed to be instantaneously released to the environment for the site boundary dose calculation.
(c) Theflashingfraction in the intact steam generator.
RESPONSE: 125,250 lbm of steam is released from each steam generator. The entire initial activity from each steam generator is assumed to be instantaneously released to the environment for the site boundary dose calculation. Thus, the flashing fraction is conservatively assumed to be 100%.
(d) The scrubbingfraction in the intact steam generator.
RESPONSE: No scrubbing is assumed.
(e) The time to isolate thefaulted steam generator.
RESPONSE: The main steam isolation valves close in 70.1 sec.
(f) The duration of the plant cooldown by the secondary side.
RESPONSE: The two-hour off-site dose was determined by releasing the entire steam l generator inventory and assuming the Technical Specification limit for the primary-to-secondary leak rate over a two hour period. Therefore, it was not necessary to determine the duration of the plant cooldown.
- 2. For the Steam 'ienerator Tube Rupture Event, provide:
(a) Tic mass ofliquid andsteam releasedfrom thefaulted steam generator.
RESPONSE: The mass of steam released from the faulted steam generator as a function of time is shown in Figures 1 and 2.
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. ATTACHMENT (2) !
, , . RESPONSE TO ADDITIONAL NRC QUESTIONS (As Provided by March 12, 1997 Telephone Conference) l 1
(b) The mass of steam releasedfrom the intact steam generator as afunction of tiene. (As a minimum. releases should be designated as those within two hours, and those after two hours.) l RESPONSE: The mass of steam released from the intact steam generator is not provided since the steam released would account for only 0.1% to 0.2% of the total releases; there are no liquid releases.
(c) Theflashingfraction in the intact andfaulted steam generators.
RESPONSE: The flashing fraction for the faulted steam generator is shown in Figure 3.
(d) The scrubbingfraction in the intact andfaultedsteam generators.
RESPONSE: The scrubbing fraction (partition coefficient) for the faulted steam generator is 0.01.
(e) The time to isolate thefaultedsteam generator.
RESPONSE: The time to isolate the faulted steam generator is 3,701 seconds. l (f) The duration ofthe plant cooldown by the secondary side. I l
RESPONSE: The plant cooldown for the first eight hours is shown in Figure 4.
(g) Primary-to-secondary release ratefrom the ruptured tube as afunction of time.
RESPONSE
- The integrated primary-to-secondary releases from the ruptured tube are shown in Figure 5.
(f) Whether over-filed conditions exist. If they do exist, appropriate mass release data should be '
provided as afunction of timefor thefaulted steam generator.
RESPONSE: The steam generator does not overfill.
- 3. For the Seized Rotor Event, provide:
(a) Liquid releasedfrom the steam generators as afunction oftime. 1 RESPONSE: 204,500 lbm is released during the first 1800 seconds, while the release rate for ,
times greater than 1800 seconds is 77.216 lbm/sec. Thus, for a cooldown period I of 8,647.2 seconds, another 667,700 lbm of steam is released, for a total of 872,200 lbm.
(b) The duration ofplant cooldown by secondary side.
RESPONSE: It requires 1800 seconds for the Reactor Coolant System (RCS) to reach 540.2 F l at 2333 psia by steaming through the Main Steam Safety Valves. Cooldown to 300 F via steaming to the condenser requires another 8647.2 seconds at l 100 F/hr. !
(c) A description of how primary-to-secondary releases were modeled as releases to the environment. l RESPONSE: Primary-to-secondary leakage of the 5% failed fuel and of the Techmca. 3 Specification RCS activity was set at the Technical Specification value of j 100 gpd per steam generator at 70*F. For the first 1800 seconds of steammg l l
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, MACIIMENT (2)
, , RESPONSE TO ADDITIONAL NRC QUESTIONS (As Provided by March 12, 1997 Telephone Conference) through the Main Steam Safety Valves, it is assumed that the RCS activity that leaks into the steam generators goes directly and completely to the atmosphere.
After 1800 seconds, a partition factor of 0.0005 is applied to the RCS activity that leaks into the steam generator and is then directed to the condenser.
- 4. For the CEA Ejection Event, provide:
(a) Thefraction offuel rods which have cladding breached as a result ofthe accident.
REscoNSE: The fraction of fuel rods which have cladding breached as a result of the accident is 10%.
(b) Thefraction offuel rods which reach or exceed initiation temperature forfuel melting as a result ofthis accident.
RESPONSE: The fraction of fuel rods which reach or exceed initiation temperature for fuel melting as a result of this accident is 10%.
(c) For the release via the primary-to-secondary pathway, a description of the assumptions utili:ed in the release of such activity. A description of how primary to secondary releases were modeled as releases to the environment.
RESPDNSE: Release via the primary-to-secondary pathway is not considered, based on the reduction and reversal of the primary-to-secondary differential pressure during the accident.
REFERENCES:
(1) Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis Report, Revision 20 (2) Letter from Mr. J. A. Tiernan (BGE) to NRC Document Control Desk, dated March 27,1987, Unit 2 Cycle 8 Reload - Request for Additional Information i
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. ATTACHMENT (2) i , ,
RESPONSE TO ADDITIONAL NRC QUESTIONS i (As Provided by March 12, 1997 Telephone Conference)
FIGURE 1 i
STEAM GENERATOR TUBE RUPTURE WITH EOP-BASED OPERATOR ACTIONS STEAM GENERATOR SAFETY VALVE FLOW VS. TIME 1
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_, MCHMENT (2) 1 E
, , RESPONSE TO ADDITIONAL NRC QUESTIONS
- (As Provided by March 12, 1997 Telephone Conference) i FIGURE 2 STEAM GENERATOR TUBE RUPTURE i WITH EOP-BASED OPERATOR ACTIONS i
! INTEGRATED ATMOSPHERIC DUMP VALVE FLOW VS. TIME i 60 i
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. - . . . . . . . . . . . . - . . .- - . _ . - . . . . . . . . ~ . . - - . - - . - . - . - _ . - _ . _ _ - . . . . .
ATTACHMENT (2) l RESPONSE TO ADDITIONAL NRC QUESTIONS (As Provided by March 12, 1997 Telephone Conference) !
FIGURE 3 i STEAM GENERATOR TUBE RUPTURE WITH EOP-BASED OPERATOR ACTIONS FRACTION OF LEAK FLASHED VS. TIME 1
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0 1 2 3 4 5 6 7~ 8 Thousands Time, seconds 6
, ATTACHMENT (2)
, , RESPONSE TO ADDITIONAL NRC QUESTIONS
- (As Provided by March 12, 1997 Telephone Conference)
FIGURE 4
- STEAM GENERATOR TUBE RUPTURE l WITII EOP-BASED OPERATOR ACTIONS CORE COOLANT TEMPERATURE VS. TIME (0 - 8 HRS.)
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y 500 -
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l, ATTACHMENT (2)
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- , , RESPONSE TO ADDITIONAL NRC QUESTIONS j (As Provided by March 12, 1997 Telephone Conference) i i
FIGURE 5
- STEAM GENERATOR TUBE RUPTURE WITH EOP-BASED OPERATOR ACTIONS k INTEGRATED LEAK FLOW VS. TIME 4
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O 1 2. 3 4 5 6 7 8 Thousands Time, seconds 8