3F0198-15, Provides Response to RAI Dtd 971209,re LAR 218 Concerning Rev of Makeup Sys Letdown Line Failure Accident Analysis

Provides Response to RAI Dtd 971209,re LAR 218 Concerning Rev of Makeup Sys Letdown Line Failure Accident Analysis

ML20198T532
Person / Time
Site:	Crystal River
Issue date:	01/20/1998
From:	Rencheck M FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0198-15, 3F198-15, TAC-M99571, NUDOCS 9801270085
Download: ML20198T532 (8)

Text

_ - - - - _ _ - - - - _ - _ - _ _ - - .

c Florida Powero" -

2 " mo

• "dM*E.".k o en-n January 20,1998 3F019815 U.S. Nuclear Regulatory Commission .

Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Response to Request for Additional Information Regarding License Amendment Request #218 - Revision of the Makeup System Letdown Line Failure Accident Analysis (TAC No. M99571)

Reference:

NRC to FPC letter, 3N1297-07, dated December 9,1997, " Crystal River Unit 3 (CR3)- Request for Additional Information - License Amendment Related to Revised Analysis of Makeup System letdown Line Failure Accident (TAC No. .

M99571)"

Dear Sir:

This submittal provides Florida Power Corporation's (FPC) response to the request for additional information dated December 9,1997, regarding License Amendment Request (LAR) #218 for Crystal River Unit 3 (CR-3). The response is provided in the attachment.

The information provided in this response does not alter any conclusions of LAR #218. 5 There are no commitments in this letter, if you have any questions regarding this submittal, please contact Mr. David Kuasemiller, Manager, Nuclear Licensir.g at (352) 563-4566.

Sincerely, t

Y f> l_ -

M. W. Rench ek, Director Nuclear Engineering and Projects M W R:dah Attachment l

xc: Regional Administrator, Region 11 l NRR Project Manager p i; p lI 0 1l- lj

' Senior Resident inspector ,1; ,I! ,.i , , 111 1, :1 gD b OON $$$302 P PDft CRYSTAL RIVER ENERGY COMPLEX 15760 W Power une Street

• Crystal River. Flonda 34428 4708 * (352) 795 4486 1 A Flonda Progres2 Compatsy

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

U. S. Nucle:r Regulatory Commission Attachment l- 3F019815 Page 1 of 7 p ATTACllMENT RESPONSC TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LAR #218 - REVISION OF Tile MAKEUP SYSTEM LETDOWN LINE 2.slLURE ACCIDENT ANALYSIS introduction License Amendment Request (LAR) #218, " Revision of the Makeup System Letdown Line Failure Accident Analysis," (Reference 7) presented a revised analysis for the hypothetical letdown line failure as part of a license amendment request for the addition of a manual operator action to terminate the analysis. The offsite radiological dose results of this analysis were also recalculated and provided in the LAR. Adciitionally, LAR #218 clearly states this is a hypothetical accident because a break in the high energy portion of the makeup system letdown line outside containment is not a credible event, llowever, this accident is analyzed to provide a bounding analysis of offsite dose consequences for lines carrying reactor coolant outside the reactor building.

Tne following provides the responses to the request for additional information receiv< d in ]

Reference 12.

Erspense to Request for Additional Information A. NRC Rcquest l'our re-analysis assumed the release of design reactor coolant activity based on 1%

failed fuct and corresponding to a dose equivalent lodine-131 activity of about 6 microcuries per gram. This is not consistent with CR3 Technical Specyications (TS) 3.4.15.A.1, which allows operation with much higher reactor coolant activity levels of 60 microcuries per gram. Our analysis, based upon CR3 TS reactor coolant activity determined the thyroid dose to be 39 rem. Our basisfor the analysis is provided in the enclosure to this letter.

Response

The radiological release calculated for the makeup system ledown line failure accident is an offsite dose analysis and did not evaluate the doses to the control room operator.

The offs'te dose is based upon one percent defective fuel rods consistent with the licensing basis for CR 3. The assumption of one percent defective fuel rods is conservative when compared to the Technical Specifidion operating limit of s;1.0 pCi/gm for Dose Equivalent lodine (1) -131 in the reactor coolant. The reactor coolant activity corresponding with a value of one percent defective fuel rods is approximately 7 Cilgm Dose Equivalent 1-131. The reactor coolant activity of 60 pCi/gm used in the NRC analysis is inconsistent with the CR-3 licensing basis and results in unrealistically high calculated thyroid doses to the control room operators. Use of the correct value for l

i

U. S. Nuclear Regulatory Commission Attachment 3F019815 Page 2 of 7 reactor coolant activity would demonstrate that the design meets the acceptance criterion of General Design Criterion (GDC) 19. Funher discussion of these items follows:

CR-3 Licensingjasis Original Analysis AS stated in CR-3 Final Safety Analysis Report (FSAR) Section 14.2.2.6.3, offsite dose calculations for the makeup system letdown line failure accident are based upon the fission product concentrations corresponding to one percent defective fuel rods.

This reflects the CR-3 licensing basis for this accident analysis. The corresponding reactor coolant activities fer he fission product concentrations for Cycle 11 were provided in revised FSAR Table 14-41 in LAR #218.

The Makeup System Ixtdown Line Failure Accident was initially analyzed (bliowing the guidance in Regulatory Guide 1.70, Revision 3, in BAW-1521. " Crystal River Unit 3 - Cycle 2 Reload Report," which was provided to the NRC in Reference 1.

BAW 1521 was reviewed by the NRC in the Safety Evaluation Report (SER) approving Amendment No.19 to the Technical Specifications (Reference 2). The Makeup System Letdown Line Failure Accident analysis was incorporated into the FSAR in 1982 in Revision 0 of the updated FSAR.

High Energy Line Break Analysis FPC has stated that the makeup system letdown line failure is not a credible break based on the methods and general criteria used to postulate and protect pipe rupture effects outside the reactor building at CR-3 as describcd in the report " Pipe Rupture Analysis Criteria Outside the Reactor Building Crystal River Unit 3." This report was submitted to the NRC in References 1 and 5. The NRC reviewed this report and found it acceptable as documented in Reference 6.

The pipe rupture report concluded that the high energy portion of die letdown line outside containment is not subject to a high energy line break. The NRC Standard Review Plan Section 3.6.2 (Reference 15) allows the establishment of a "No Break Zone" if certain criteria are met. Also, per Generic Ixtter 87-11, " Relaxation in Arbitrary Intermediate Pipe Rupture Requirements," (Reference 16) arbitrary intermediate piping breaks need not be postulated if certain criteria are met.

Revision 1 of the Pipe Rupture Analysis established the "No Break Zone" (NBZ) for the letuown line piping between the containment isolation valves for this specific containment penetration (Reference 5). The NBZ was based upon the methods specified in the Standard Review Plan. Additionally, the report determined that the stress in the piping from the containment to the manual : solation valves downstream of the block orifice and letdown control valves is low enough that an arbitrary intermediate break need not be considered in this section of the letdown line.

Therefore, a break in the high energy portion of the letdown line omiide containment is not considered a credible event. '

U. S. Nuclear Regulatory Commission Attachment 3*c0198-15 Page 3 of 7 The NRC acceptance of Revirion 1 of the Pipe Rupture Analysis in Reference 6 stated, "We have reviewed the revised sections of the submitted report and have determined that the revised sections incorporate the statements, conditions, and criteria required by our letters of September 28 and November 9,1989, and are acceptable." Based on the requirements of the Standard Review Plan and Generic Letter 87-11, designing for the dynamic or environmental effects of a high energy line break in the letdown line outside containment is not required.

Current Analysis Therefore, the Makeup System letdown Line Failure Accident is presented only to demonstrate that the offsite dose consequences from a postulated break in the letdown line outside containment remain below the 10 CFR 100 limits. Relative to offsite dose consequences, the hypothetical break in the letdown line bounds other postulated breaks in lines that carry reactor coolant outside containment.

Technical Specification Bequirements Liraiting Conditionsfor Operation (LCO) and Action Statements 3

FPC safety analyses do not assume design basis accidents are initiated when in an action statement allowed by the Technical Specifications. The CR-3 operational limit for D9se Equivalent I-131 is 1.0 pCi/gm as specified in Technical Specification 3.4.15. If the Dose Equivalent 1-131 exceeds 1.0 pCi/gm, the appropriate Action statement must be entered. Technical Specification Figure 1 d 15-1, which shows a limit of 60 pCi/gm between 80 and 100 percent rated u.nal power, does not

' represent the operational limit, but the limit associated with Required Action A.1 of Technical Specification 3.4.15.

Operation with the Dose Equivalent 1-131 greater than 1.0 pCilgm is limited by the Action requirements to less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if the activity is within the values of Technical Specification Figure 3.4.15-1. If the Dose Equivalent I-131 is not restored to s 1.0 pCilgm within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or the activity is not within the values of Technical Specification Figure 3.4.15-1, the plant must be placed in Mode 3 with T,,< 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Therefore, the operating time with > 1.0 pCi/gm of Dose o Equivalent 1-131 is limited.

' Whenever the plant enters an Action statement, it is recognized that full two-train r operability performance of the effected system may have been Jegraded to the point y where only one train operation may be all that is available to this system. In fact, ITS Section 1.3, " Completion Times," provides two such examples where an Emergency Core Cooling System (ECCS) may become impaired such that two separate functions, one in each of two separate trains of an ECCS, can become 3 inoperable or a ccmplete train can become inoperable and the plant is allowed to continue to operate for a limited Action statement duration.

a l t

'U. S. Nucle:r Regulatory Commission Attachment 3F019815 Page 4 of 7 Bases The Technical Specification Bases for 3.4.15 state that limited operation with Dose Equivalent 1-131 greater than 1 pCi/gm is allowed by Action A to provide a reasonable time for temporary coolant activity increases to be cleaned up with processing systems. As stated in the Technical Specification Bases for 3.4.15, the value of 1.0 pCi/gm is the operating limit and represents a reasonable operating capability rather than a specific analytical result. The Bases specifically discuss the

r. team generator tube rupture (SGTR) analysis as assuming the specific activity of the reactor coolant is representative of one percent defective fuel.

The Technical Specification Bases for 3.4.15. " Applicable Safety Analysis," further states that "RCS Specific Activity satisfies Criterion 2 of the NRC Policy Statement." This refers to 10 CFR 50.36(c)(2), " Limiting conditions for operation,"

(ii)(B):

C,iterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the intagrity of a fission product barrier.

This defines that the pulpe of this LCO is to establish the operating restriction for RCS Specific Activity that is the expected initial condition for accident analyses.

The LCO limit in this case is 1 pci/gm which is lower than the value used in the SGTR accident analyses (1 % failed fuel). The same initial conditions should be used for other analyses as well.

The LCO, Actions, and Bases in CR-3 Technical Specification 3.4.15 are consistent with the Babcock and Wilcox (B&W) Standard Technical Specifications in NUREG-1423.

B. NRC Request Please submit the control room dose calculations that demonstrate CR3 satisfies the requirements of GDC 19.

Response

NUREG-0737, Item Ill.D.3.4, Control Room liabitability Requirements," requires that licensees assure that control room operators will be adequately protected against the effects of accidental release of radioactive material and that the nuclear power plant can be safely operated or shut down under design basis accident conditions. In Reference 4, the NRC concluded that the CR-3 control room habitability design met the criteria identified in NUREG-0737, Item Ill.D.3.4 and was therefore acceptable. As stated at that time, the Maximum Hypothetical Accident (MHA), which consists of a loss of coolant accident (LOCA) concurrent with a loss of offsite power (LOOP), was the bounding design basis event for the control room radiological evaluation.

e

j ,

U. S. Nuclear Regulatory Commission Attachment 3F0198-15 Page 5 of 7 During the current outage, the control room habitability analyses have been reperfonned using a different calculational methodology from what was previously used to demonstrate compliance with NUREG-0737, item Ill.D.3.4 (Reference 10). The results of these analyses have been provided as Attachment D of LAR #222 (Reference 11).

Based on the evaluations and assessments perfonned, the MilA with a LOOP was detennined to be the limiting accident for control room habitability. The results of ti.is analysis demonstrates that the limits of GDC 19 are not exceeded.

In determining the limiting event for control room habitability, a review of other design hasis accidents for CR-3 was performed to verify that the MilA was the limiting event.

As part of this assessment, the SGTR accident was evaluated for those events where reliance on the radiation monitor or operator action to isolate the control room emergency ventilation system (CREVS) is required. This analysis demonstrated that isolation of the CREVS was not necessary to maintain operator exposures less than regulatory limits. Further, given any reasonable isolation time either by the radiation monitor or operator action, the MilA remains the bounding event with regard to control room habitability.

In Reference 9, FPC responded to the NRC request for additional infonnation (Reference

8) regarding the radiological dose to the control room 4., ms -s a result of the makeup system letdown line failure accident. As stated in iponse, the m;keup system letdown line failure accident is bounded by the Mita for dose to the control room operators.

Compared to other accidents involving radiologietd releases, the dose to the control wom operators associated with the MilA is more bounding than the design basis SGTR or the makeup system letdown line failure accident. The offsite dose consequences of the SGTR are either bounded by, or consistent with, those observed for the makeup system letdown line failure accident, with auxiliary building filtration. The offsite dose:, for the makeup system letdown line failure accident are also calculated with no credit for auxiliary building filtration. With no credit for filtration, the makeup system letdown line failure accident causes higher offsite dose consequences. Regardless, both are events which demonstrate considerable margin to the 10 CFR 100 limits. Moreover, it had been previously detennined that the control room dose due to an SGTR has been found to be considerably less than those calculated for the MilA. Further, although no specific analysis for control room habitability was perfonned for the makeup system letdown line failure accident, an esaluation based on the activity released to the environment indicates that the control room dose consequences are also less than the MilA. In conclusion, neither accident can result in consequences of the severity calculated for the :,tilA, and the MllA remains the limiting accident with respect to both offsite and control room operator doses.

The control room dose calculations for CR-3 were discussed with the NRC in a meeting on January 8,1998. Agreements that were reached regarding the control room habitability report and the corresponding calculations are detailed in Reference 14.

Therefore, the operator dose calculations are not being forwarded with this submittal.

U. S. Nuclear Regulatory Commission Attachment 3F019815 Page 6 of 7 References

1. FPC to NRC letter, 3F0279-10, dated February 28,1979, " Environmental Impact Appraisal and Balance of Plant Review for Cycle 2 Reload and Power Ixvel Upgrade"

2. NRC to FPC letter,3N0779-01, dated July 3,1979, " Amendment No.19"

3. FPC to NRC letter, 3F038919, dated March 31,1989. "lligh Energy Line Break Outside Reactor Building Criteria" ,

4. NRC to FPC letter,3N0589-25, dated May 25,1989, " Crystal River Unit 3 - Control Room IIabitability Evaluation (NUREG-0737 Item III.D.3.4) (TAC NO. 64805)"

5. FPC to NRC letter, 3F1289-11, December 18,1989, "lligh Energy Line Break (IIELB), Revision cf Design Criteria and Schedule Update"

6. NRC to FPC letter,3N0490-10, dated April 11,1990, " Crystal River Unit 3 - liigh Energy Line Break (IIELB) Criteria for Analysis of Piping Outside Containment (TAC No. 69533)"

7. FPC to NRC letter, 3F0997-14, dated September 9,1997, " License Amendment Request 218 - Revision of the Makeup System Extdown Line Failure Accident Analysis"

8. NRC to FPC letter, 3N1097-13, dated October 10,1997, " Crystal River Unit 3 -

Request for Additional Information - License Amendment Related to R.: vised Analysis of Makeup System 1xtdown L.ine Failure Accident (TAC No. M99571)"

9, FPC to NRC letter, 3F1197-26, dated November 7,1997, " Response to Request for Additional Information - License Amendment Request 218, Revision of the Makeup System 1.ctdown Line Failure Accident Analysis (TAC No. M99571)"

10. FPC to NRC letter, 3F1197-09, dated November 10,1997 " Control Room Itabitability, NUREG-0737, item Ill.D.3.4"

11. FPC to NRC letter, 3F1297-19, dated December 5,1997, " License Amendment Request #222, Revision 0 - Control Room Emergency Ventilation and Emergency Filters"

12. - NRC to FPC letter, 3N1297-07, dated December 9,1997, " Crystal River Unit 3 (CR3) - Request for Additional Information - License Amendment Related to Revised Analysis of Makeup System Letdown Line Failure Accident (TAC No. M99571)"

13. NRC to FPC letter, 3N1297-19, dated December 24,1997, " Crystal River Nuclear Generating Plant Unit 3 - Control Complex Ilabitability Envelope Justification for Contined Operation - Interim Assessment Results (TAC M91823)"

U. S. Nuclear Regulatory Commission Attachment l

j 3F0198-15 Page 7 of 7

14. FPC to NRC letter, 3F0198 26, dated January 14,1998, " Control Complex liabitability Envelope Justification for Continued Operation (TAC M91823)"

15. NUREG-0800 dated July 1981, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition"

16. Generic 1. citer 87-11 dated June 19,1987, " Relaxation in Arbitrary Intermediate Pipe Rupture Requirements" i

-- - - - - - - . . _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _____