ML20209G559

Forwards Detailed Listing of Staff Concerns with C-E TR NPSD-683,rev 3 & RCS pressure-temperature Limits Rept

ML20209G559
Person / Time
Site:	Fort Calhoun
Issue date:	07/08/1999
From:	Wharton L NRC (Affiliation Not Assigned)
To:	Gambhir S OMAHA PUBLIC POWER DISTRICT
References
TAC-MA5537, NUDOCS 9907190195
Download: ML20209G559 (15)

Text

...

July 8, 1999 j 3-

.. Mr. S. K. Gambhir -

Division Manager - Nuclear Operations Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.' '

Post Office Box 399 Hwy. 75 - North of Fort Calhoun Fort Calhoun, NE 68023-0399:

I

SUBJECT:

CONCERNS WITH PRESSURE TEMPERATURE LIMITS REPORT (PTLR) AND j TOPICAL REPORT CE NPSD-683 (TAC NO. MA5537) l

Dear Mr. Gambhir:

By [[letter::LIC-99-0045, Forwards Application for Amend to License DPR-40,proposing to Relocate pressure-temp Curves,Predicated NDTT Shift Curve & LTOP Limits & Values from FCS TS to OPPD Controlled Document.Rev 3 to TR CE NPSD-683 & Supporting TSs Encl|letter dated May 26,1999]], Omaha Public Power District (OPPD) submitted the Pressure Temperature Limits Report (PTLR) amendment request and Combustion Engineering Owners Group (CEOG) Topical Report CE NPSD-683, Revision 3 for staff review. The staff has performed an acceptance review and identified concerns with the submittal. Some of these concems have been discussed with you by telephone on June 17,1999. In further discussio,1s with you on June 28,1999, you decided to withdraw this submittal. As a follow-up to these discussions, we have enclosed a detailed listing of staff concerns related to the submittal for your consideration.

Sincerely, )

L. nard Wharton, Project Manager, Section 2 Project Directorate' IV & Decommissioning

- Division of Licensing Project Managemc...

, . Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosure:

Staff Concerns '

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% . . . . . #o Mr. S. K. Gambhir Division Manager - Nuclear Operations Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

Post Office Box 399 Hwy. 75 - North of Fort Calhoun Fort Calhoun, NE 68023-0399

SUBJECT:

CONCERNS WITH PRESSURE TEMPERATURE LIMITS REPORT (PTLR) AND TOPICAL REPORT CE NPSD-683 (TAC NO. MA5537)

Dear Mr. Gambhir By [[letter::LIC-99-0045, Forwards Application for Amend to License DPR-40,proposing to Relocate pressure-temp Curves,Predicated NDTT Shift Curve & LTOP Limits & Values from FCS TS to OPPD Controlled Document.Rev 3 to TR CE NPSD-683 & Supporting TSs Encl|letter dated May 26,1999]], Omaha Public Power District (OPPD) submitted the Pressure Temperature Limits Report (PTLR) amendment request and Combustion Engineering

)

Owners Group (CEOG) Topical Report CE NPSD-683, Revision 3 for staff review. The staff has performed an acceptance review and identified concems with the submittal. Some of these concerns have been discussed with you by telephone on June 17,1999. In further discussions with you on June 28,1999, you decided to withdraw this submittal. As a follow-up to these discussions, we have enclosed a detailed listing of staff concerns related to the submittal for your consideration.

Sincerely, 1

I L. Raynard Wharton, Project Manager, Section 2 Project Directorate IV & Decommissioning Division of Licensing Project Management l Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosure:

Staff Concerns 1

cc w/ encl: See next page l

Ft. Calhoun Station, Unit 1 cc:

Winston & Strawn ATTN: Perry D. Robinson, Esq.

1400 L Street, N.W.

Washington, DC 20005-3502 i

Mr. Jack Jensen, Chairman Washington County Boani of Supervisors 1 Blair, Nebraska 68008 -

Mr. Wayne Walker, Resident inspector U.S. Nuclear Regulatory Commission j Post Office Box 309 l Fort Calhoun, Nebraska 68023 Regional Administrator, Region IV q U.S. Nuclear Regulatory Commission l 611 Ryan Plaza Drive, Guite 1000 }

Arlington, Texas 76011

(

Ms. Cheryl Rodgers, LLRW Program Manager j Environmental Protection Section Nebraska Department of Health 301 Centennial Mall, South P.O. Box 95007 Lincoln, Nebraska 68509-5007' Mr. J. M. Solymossy Manager- Fort Calhoun Station Omaha Public Power District Fort Calhoun Station FC-1-1 Plant Post Office Box 399

. Hwy. 75 - North of Fort Calhoun ,

Fort Calhoun, Nebraska 68023 Mr. Mark T. Frans Manager- Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

Post Office Box 399 Hwy. 75 - North of Fort Calhoun Fort Calhoun, Nebraska 68023-0399

I

\

I STAFF CONCERNS WITH COMBUSTION ENGINEERING TOPICAL REPORT CE NPSD-683. REV. 3 AND RCS PRESSURE-TEMPERATURE LIMITS REPORT (PTLR)

OMAHA PUBLIC POWER DISTRICT l

FORT CALHOUN STATION. UNIT 1 l DOCKET NO. 50-285 L CONCERNS WITH COMBUSTION ENGINEERING TOPICAL REPORT CE NPSD-683,

, REV. 3 A. Chapter A-Introduction On page 6, the first paragraph of the page states that once the' pressure temperature limits report (PTLR) has been incorporated into the plant's technical specifications, any changes made in a PTLR would be controlled by the requirements of 10 CFR 50.59 and that the changes would no longer require a license amendment submittal to become effective. We have informed CE and OPPD that a plant could make changes in th(

PTLR through the design change (10 CFR 50.59) process if the changes in the P-T limits or LTOP setpoints were calculated using the approved methodology. However, the PTLR process requires a licensee to submit a new administrative section that refers

. to the specific version of the methodology that has been approved by the staff for generating P-T limit curves and LTOP system setpoints. The staff has previously emphasized that if a licensee was proposing a change to the approved methodology, the licensee would have to submit a license amendment request; this is consistent with {

i the staff's position on page 2 of Generic Letter 96-03. Changes to the approved methodologies for the P-T limit curves and for the LTOP settings cannot be accomplished through the 10 CFR 50.59 process. The paragraph needs to be reworded j to reflect this.

B.' Chapter 1.0 - Neutron Fluence Calculation Methods (Pages 13-26)

1. The fluence calculation proposed in Section 1.0 of this report does not constitute a

" Methodology" but it is acceptable for plant specific applications. The staff assumes 4 that Section 1.4.1 does not include calculational adjustments based on plant specific data.

OPPD, Fort Calhoun PTLR Submittal.

. Refs. (1) LIC-98-0009 dated January 30,1998, Attachment 3, and (2) LIC-99-0045 dated May 26,1999, Attachment C, Section 2.1.

ENCLOSURE

2. The value of 1.15x10 n/cm' is cited from Ref.1 but the value of 1.501x10" n/cm2 is recommended. Neither of the above values is acceptable. The 1.15x10" n/cm r has been derived using plant specific data adjustments which is in violation of the .

method proposed in CE NPSD-683 Rev. 3. The 1.501x10" n/cm2 value was derived

- using ENDF/B-IV based cross sections which is not conservative particularly for a thermal shield plant. On the other hand low leakage loading stratogies have been applied for several Ft. Calhoun cycles which is conservative. However, the submittal does not quantify these effects to justify why the recommended value is .

conservative.

C. Chapter 2.0 - Reactor Vessel Surveillance Program (Pages 27-29)

The chapter states on the bottom of page 28 that a proposed modification to the surveillance capsule withdrawal schedule can be evaluated under the provisions of the 10 CFR 50.59 process if the withdrawal (removal) schedules are not specified in the Technical Specifications. Part 50, Appendix H, Section Ill.B.1 of 10 CFR states that the design of the surveillance capsule programs and withdrawal schedules must meet the requirements of the Edition of ASTM Standard Procedure E-185 which is current on the issue date of the ASME Code to which the reactor vessel was purchased. Section Ill.B.1 of the Appendix also states that later editions of ASTM E-185 may be used through the 1982 edition of the Standard Procedure. The staff position is that a licensee can use the 10 CFR 50.59 process to amend a previously approved surveillance capsule withdrawal schedule, only if the withdrawal schedule was not located in the plant's Technical Specifications, and if the proposed changes were consistent with the licensee's ASTM E-185 procedure of record, or with one of the more recent editions of ,

the Standard Procedure listed in the rule (e.g., ASTM Standard Procedures E185-73, E185-79, or E185-82). Otherwise, pursuant to 10 CFR Part 50, Appendix H, Section Ill.B.3., such proposed changes would have to be submitted for review and approval of the staff. As stated on page 28 of the report, if the surveillance capsule withdrawal schedule is located in the Technical Specifications, any proposed changes to the schedule would require a license amendment request submittal (pursuant to 10 CFR 50.90 submittal). The report needs to be revised to reflect these requirements and the restrictions on using the 10 CFR 50.59 process for changes to the withdrawal schedule.

_ D. Chapter 3.0 - Low Temperature Overpressure Protection The staff has identified a number of issues that need to be addressed regarding the generic topical report. Most of there issues are relatively simple, however, a number of significant issues remain. The significant issues are: the continued ambiguity regerding the appropriate references that can be used to develop the P-T limits and the LTOP P-T limits, the lack of a model for calculating the energy addition transient, ambiguity regarding how a steam bubble in the pressurizer will be credited in the analysis, ambiguity regarding how operating restrictions can be credited in the analysis and vague statements regarding what assumptions need to be included in the analysis. A list of all  !

I the concerns is provided below.

I

1. The discussion in Section 3.0 regarding the development of the P-T limits and LTOP limits is very general, confusing and could be misleading. Statements implying that a 1.1 relaxation factor can be applied to references 10 or 11 must be removed.

The topical should state clearly that there are only two acceptable methods for generating P-T limits: (1) Appendix G to Section XI of the 1986 ASME code, or (2)

ASME Code Case N-640, if an exemption is granted by the NRC. Additionally, the topical should state clearly that there are only three acceptable methods for generating the LTOP P-T limits: (1) Appendix G to Section XI of the 1986 ASME code, (2) 110% of Appendix G to Section XI of the 1986 ASME code, if an exemption is granted from the NRC, or (3) ASME Code Case N-640, if an exemption

, is granted from the NRC.

Although the topical can be interpreted a number of ways, the above is how the staff have interpreted the discussion. If this is not the case, state clearly what else would be acceptable under this topical and why. I l

2. The general methodology is based on the presumption that an adequate LTOP system can be designed a number of ways by varying the assumptions. The methodology implements this strategy by allowing a number of plant parameters to ,

be controlled in the PTLR, rather than in the TS. Although this concept could be appealing it goes well beyond the process described in Generic Letter 96-03.

Additionally, generic TS changes and reductions in TS content are being discussed with the Owners Groups. As a result, it is not appropriate to move these controls to the PTLR.

3. The statement on page 38 of the topical, that operating restrictions that reduce the severity or eliminate a transient "shall be placed in the TS" is misleading. The analyses need to be based on the TS. The topical should state that if there are no TS controlling the restriction, then the restriction cannot be credited in the analysis or put in the PTLR. For example, for plants without a TS on the charging pumps, pressurizer level in modes 4, 5 and 6, or reactor coolant pumps in operation, the topical should state that these restrictions cannot be credited in the analysis or put in the PTLR.

4. Please clarify how a pressurizer steam volume is used as an additional qualifier in the overpressure analyses? The discussion implies that it be used in lieu of a requirement on the relief valves rather than in conjunction with the relief valves for both the mass addition and energy addition transient? The steam bubble in conjunction with operator action should be considered addibonal defense-in-depth or margin when performing the water-solid calculations. For example, it is prudent to assure there is a steam bubble prior to starting a RCP to prevent the relief valves from being challenged, however, the overpressure analysis should generally

- consider water-sol,J conditions. If an individual plant needs this credit as a result of having only one relief valve or the plant's design basis already credits the steam bubble, this can be credited, however, generic approval is not appropriate.

5. The topical should indicate that pressurizer level uncertainties need to be considered in the analysis and indicate which standard should be used for determining the uncertainties.

6. On page 36, with respect to the mass addition transient, the topical states the limiting event is the simultaneous operation of two HPSI and three charging pumps or the combination of the maximum flowrate permitted by TS. The plant specific discussion of two HPSI or three charging pumps should ba removed or used as an

' example. The maximum flowrate permitted by TS should be the only criteria.

7. To reduce confusion please define " conservative margin" when evaluating pump performance ;page 45).

8. The methodology needs to require that the core flood tank pressure be verified to determine if they need to be isolated. ,

9. Pages 38 and 39 discuss the assumptions for the analysis, however, it includes a number of statements, "unless a less restrictive approach is justified." These statements should be removed. It is not clear who needs to justify the altemative approach, the NRC, the licensee, or the vendor. Does changing these assumptions mean the methodology is not being followed? If these assumptions are not considered part of the methodology, why are they not?

10. Wdh regard to the " typically used" assumptions on page 44, please describe why these are not considered important and part of the methodology. Additionally, how i is the steam generator heat transfer surface area for the energy addition transient determined.

11. For the operating and discharge characteristics of the SDC relief valves, please J include the statement that the ASME standards and/or manufacturers I recommendations, "whichever is more conservative," should be used. Additionally, j the inlet pressure drop should also be included for these valves (if not included in the discharge characteristics).

12. With regard to the pressure difference between the pressurizer and the limiting weld accounted for in the P-T limits and setpoints, the topical should state that maximum number of RCPs and RHR pumps, permitted by TS, should be accounted for in the P-T limits unless there is a TS restnction on RCP operaten in modn,s 4, 5, and 6.

13. With regard to operator action, within "10 minutes of the start." Please clarify that if credit for operator action given it should be assumed 10 minutes after being alerted to the problem, not 10 minutes from the start of the event.

14. On page 42 the report states that a pressure vs setpoint function can be generated.

Please describe how this is developed and how it will be used.

15. The anergy addition evaluation method or analytical model for this event is not provided. A description of this model needs to be provided or the topical needs to state that a separate NRC approved model is needed and will be referenced in the TS administrative controls section to apply this PTLR methodology.

16. The ABB CENP method of equilibrium pressure method appears to be an acceptable model for water solid conditions. However, it is not clear how this model will be applied when credit is given for a pressuiizer steam volume. Please describe how these time-dependant calculations are performed.

17. There should be a sample set of marked-up TS pages for the CE Standard TS. The marked up TS pages provided does not include an Administrative Controls section.

The marked up pages need an Administrative controls section referencing the approved topical. The marked-up TS should also include TS on all restrictions credited in the report.

18. Temperature uncertainties are discussed in a number of places, however, temperature uncertainties do not seem to be considered in all applications. To clarify, please state what standard wi!! be used to quantify the temperature uncertainties and state that the uncertainties will be be applied in all cases where temperature plays a role (i.e., enable temperature, P-T limits /LTOP P-T limits or setpoints, and all cases where temperature related operating restrictions are applied).

19. For the development of the enable temperature in Section 3.4.3, how is the temperature different between the water temperature and the 1/4 or 3/4 t location calculated? Also, to clarify, an exemption is not required when the Code Case N-514 is applied only to the enable temperature, however, NRC approval is needed.

20. Please indicate in the topical that when establishing the bolt-up temperature, the P-T limits /LTOP P-T limits and setpoints that have been generated must encompass the RCS temperature associated the established bobup temperature (i.e., the P-T limits and LTOP protection bound operation with the head bolted).

21. On pages 30 and 31, the report lists and discusses what are acceptable methodologies for generating the P-T limits LTOP system setpoints. The discussion is ambiguous because it implies that any combination of methodologies for the P-T limits and the LTOP setpoints can be used in conjunction with one another.

Pursuant to the requirements of 10 CFR Part SO, Appendix G, there are restrictions on which methodologies listed for the P-T limits can be used in conjunction with the methodologies listed for the LTOP setpoints.

22. The middle paragraph of page 32 discusses how the LTOP system setpoints are established from the P-T limits. The staff considers the wording in the paragraph on page 32 to be ambiguous, in that a licensee may intepret the wording to mean that the P-T limit pressure values satisfying equation (1) of the 1996 Appendix G may be relaxed by 110% and then again by 110% to establish the LTOP system pressure setpoints.

l 1

L_ _- _

On this page, and throughout the report, the term "LTOP P-T limits" creates confusion with references to Appendix G P-T limits (e.g., P-T limits generated from the stress intensity equation in Appendix G). The 1996 edition of Appendix G does not allow the pressure values that are established from the stress intensity equat on in the App +.ndix to be multiplied by a value of 1.1. Paragraph G-2215 of the Appendix states that "LTOP systems shall limit the movimum pressure of the vessel to 110% of the pressure determined to satisfy equation (1) of Appendix G (e.g, the stress intensity equation for generating the P-T limit 9). We recommend the following: (1) throughout the report, replace the term "t. TOP P-T limits" with a terminology that avoids confusion with the Appendix G P-T limits; and (2) reword the final two sentences of the paragraph to state: "The latter requirement, which was first introduced by Reference 11, effectively increases the Appendix G P-T limits by 10% to arrive at the LTOP setpoint values. As indicated in Section 3.1.1, an exemption must be obtained from the NRC to use either Reference 10 or 11 as the basis for establishing the LTOP setpoints."

E. Chapter 4.0 - Method for Calculating Beltline Material Adjusted Reference Temperature (ART)

No issues F. Chapter 5.0 - Application of Fracture Mechanics in Constructing P-T Curves

1. The middle paragraph on page 53, the report is redundant in that it repeats the

- option of using either Code Case N514 or the 1996 Edition of Appendix G to Section XI of the Code as the besis for establishing the LTOP pressure setpoints. These options were already discussed in Section 3.0 of the report and do not need repeating. The staff's issues with the discussion of these methodologies have been been described previously in items D.21 and D.22 to this list of staff concems. To avoid confusion, the middle paragraph on page 53 needs to be deleted.

- 2.' The tcp of page' 54 discusses the acceptable methodologies for generating both the P-T limits and the LTOP setpoints and is basically a repetibon of the discussion on pago 30 and 31 of the report. Agau, the discussion is ambiguous because it imphes that any combination of methodologies for the P-T limits and the LTOP setpoints can be used in conjunction with one another. Pursuant to the -

requirements of 10 CFR 50.55a,10 CFR Part 50, Appendix G, and Appendix G of the ASME Code, there are rr4 richens on which methodologies listed for the P-T lirnits can be used in conjunction with the methodologies listed for the LTOP setpoints.

3. Page 73 lists a senes of equahons to be used in the calculations of the allowable pressure data that will be used in the generation of the P-T limit curvest. The page states that the M, factors used in the calculations may be determined froin either Figure G-2214-1 of the 1996 Edition of Appendix G to Sechon XI or from Figure G-2214-2 of one of the Pre-1996 Editions of Appendix G to Section XI. Page 73 also states the M,, factors used in the calculations may be determined from either Figure

G-2214-1 of the 1996 Edition of Appendix G to Section XI or from the corresponding M, formula la the 1996 Edition of the Appendix. At this point, the staff has only al proved Editions of the ASME Code through the 1989 Edition of the Code j (pursuant to 10 CFR 50.55a). Any reference to Figure G-2214-1 should be to the 1989 or Pre-1989 Editions of Appendix G to Section XI Furthermore, pursuant to 10 CFR Part 50, Appendix G, a licensee will need to apply for an exemption to use the 1996 Edition of Appendix G for the determination or calculation of the M. and M, coefficient values if the 1996 Edition yields higher values of the coefficients than would use of Figure G-2214-1 from one of the approved editions of the Appendix (e.g., the 1989 Edition or Pre-1989 Editions of Appendix G to Section XI).

'4. Figures 5.7 through 5.11 provide examples of tipical composite P-T limit heatup, cooldown, and hydrostatic testing P-T limit curves for CE designed nuclear plants.

At pressures greater than 20% of the preservice hydrostatic test pressure, a vertical

. line is drawn in the figures that is based on the lowest service temperature criteria (given in Section 6.3 of the report). The lowest service temperature criteria is relative to the limiting RT y value of the ferritic low alloy steel piping, pump, and valve mattabls in the primary coolant pressure boundary (e.g., set at RTer j

+100'F). However, the lowest service temperature cria.eria may be non-conservative i relative to the minimum temperature requirements for the vessel when the RCS is l pressurized to greater than 20% of the isreservice hydrostatic test pressure (PHTP). J

' It is critical to point out thei the vertical knes for pressures greater than 20% of the PHTP should be based on the criteria that yield the more conservative results. This

. issue should also be clarified in the P-T limit figures in the sample PTLR (Appendix A to the report). I G. Chapter 6.0 - Method for Addressing 10 CFR 50 Minimum Temperature Requirements in the P-T Limit Curves ,

The section only lists the minimum temperature requirements for operation with the core critical when the reactor coolant system (RCS) pressure is s 20% of the preservice hydrostatic test pressure (PHTP), and when the RCS pressure is > 20% of the PHTP.

The section does not mention the minimum temperature requirements for inservice hydrostatic / leak rate testing conditions both at s 20% and > 20% of the PHTP, and during normal operations at s 20% and > 20% of the PHTP when the core is not c.stical All of the minimum temperature requirements shou % be stated and should basically tr the same as those mentioned on pages 5940 of the report.

H. Chapter 7.0 - Application of Surveillance Capsule Data to the Calculation of Agusted Reference Temperature

1. On page 96 it is stated that a licensee using the methodology may apply the surveillance data from a sister plant that has an equivalent material (e.g., equivalent heat number) in the surveillance program for the sister plant's reactor vessel.

2. Pursuant to the requirements of 10 CFR Port 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements," the staff position is that a licensee l

l

may only use the data from a sister plant if the surveillance program and data has been approved by the staff as complying with the requirements for Integrated Surveillance Programs that are Section Ill.3.C. to 10 CFR Part 50, Appendix H. If a licensee has not been approved to use integrated surveillance data the rules require that the licensee submit a request to use the integrated data. According to the rule, such requests will be evaluated by the Director of the Office of Nuclear Reactor Regulation on a case-by-case.

1. Chapter 8.0 - Summary of Results No issues.

J. Chapter 9.0 - References j Reference 10 needs to be revised to remove the reference to the 1996 Edition of Appendix A to Section XI of the ASME Code since the report does not refer to this I methodology as being acceptable for either the establishment of the LTOP system setpoints or for generation of the P-T limit curves.

K. Appendix A to the Report 1

1. Section 2.3 of the Appendix (page A-6) states that the acceptability criterion for the LTOP system is that the " peak transient pressure does not exceed 110% of the applicable Appendix G pressure limit. Section 2.3 of the Appendix does not state that a licensee cannot apply for an exemption to set the LTOP setpoints at 110% of the peak Appendix G pressure (e.g., an exemption to use either the 1996 Edition of Appendix G to Section XI or Code Case N-514 for the LTOP pressure setpoint) if the applicant is requesting an exemption to use Code Case N-640 as the basis for calculating the Km values used in the Appendix G P-T limit calculations. The report needs to correct this omission. L l

2. Section. 2.6 of the Appendix (page A-12) should clarify that the lowest service j temperature line in Figures 4-1 and 4-2 of the Appendix should be generated from 1 the acceptance criterion that yields the more conservative value: (1) the minithum temperature requirement for normal operations with the core not critical and the

_ RCS pressure greater than 20% of the preservice hydrostatic test pressure (PHTP),

.or (2) the lowest sennce temperature requirement for the ASME Code Class 1 piping, pump and valves.

- L. Appendix 5 to the Report A

1. Appendix B providas an example of the proposed technical specifications for a j typical PTLR license amendment request from a CEOG member utility. However, -1 the sample proposed technical specifications do not contain an administrative controis technical specific.000 page which govems the PTLR program. This is not consistent with the critena in GL 96-03 and needs to be corrected.

t

2. In accordance with Generic Letter 96-03, the PTLR process requires that any changes made to a previously approved NRC methodology be submitted by the licensee to the NRC for approval. This is true for any changes to a methodology.

Changes to the curves, etc., using an approved methodology do not have to be reviewed by the NRC.

11. CONCERNS WITH FORT CALHOUN STATION PLANT-SPECIFIC APPLICATION OF CEOG PRESSURE TEMPERATURE LIMITS REPORT, MAY 26,1999 A. LTOP

1. The TS must reference a specific, dated version of the methodology. Referencing the latest approved version is not acceptable. This comment is merely confirmatory.

2. Is operator action credited in the energy addition transient to terminate the event?

Operator action with the pressurizer bubble can only be credited as one method of mitigating the event. The relief valve also needs to be verified acceptable through analysis.

3. : la the pressurizer bubble credited in the mass addition transient? If the bubble is credited in the mass addition transient, is there a TS controlling pressurizer level?

4. TS 2.1.6 should specify that the PORV setpoint is in the PTLR. The TS applicability should be based on the temperature, not based on the plant being heated up or being cooled down. TS 2.1.1 should reference the " enable temperature" in the PTLR (editorial comment).

5. An NRC approved energy addition evaluation methodology needs to be referenced in TS 5.g.6. Additionally, the definition of PTLR should reference the Administrative Controls section of the TS (TS 5.9.5) (per the standard TS).

6. The "or" in PTLR Section 2.3.1(A)(2) implies that two analysis were performed: (1) assuming 53% pressurizer volume and the maximum primary to secondary temperature difference, and (2) assuming a full pressurizer and 30*F primary to l secondary temperature difference. Is this true? If so, what was the temperature difference assumed in the first case?

7. The pressure correction factor is a function of the number of RCPs in operation. Do the TS limit the number of RCPs in operation below 210*F, and below the enable temperature. If not, why not?

B. Materials

1. Attachment 1 - the proposed TS pages

a. Basis Section of revised TS page 2-4: reference should be made to the specific staff approved version of the Generic Report (at this point CE NPSD-683, Revision 3). .

b. Reference No.10 on revised TS page 2-7a: reference should be made to the specific staff approved version of the Generic Report (at this point CE NPSD-683, Revision 3).

c. Page 5-17b of the submittal provides the proposed Administrative Controls page for the TS. The new TS Section 5.9.6, " Reactor Coolant System (RCS) Pressure

-Temperature Limits Report (PTLR), in part, states that "The analytical methods used to determine the RCS pressure and temperature limits and predicted radiation ir.duced NDTT shift shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: (1) 10 CFR 50.61, . . . . . ; (6) CE NPSD-683, " Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications, . . . . . " The staff has the following issues with proposed TS 5.9.6:

1. The nil ductility transition temperature shift is only one of a number of parameters that are used in the calculations of the adjusted reference temperature for the beltline materials in the reactor vessel; proposed TS 5.9.6 does not mention that the approved methodologies in the document list will be used for the determination of the neutron fluence values, LTOP system setpoints, adjusted reference temperatures for the reactor vessel materials and the RCS piping, pump and value materials, and the pressure-temperature limits. The proposed section should be amended to reflect this.

ii. The P-T limit curves in the PTLR all contain a lowest service temperature line. The scope of Appendix G to the 1986 Edition of Section ill applies to ferritic vessel materials, and to ferritic nozzle and flange regions attached to the vessel. Paragraph NB-2332 of the 1986 Edition of Section til to the ASME Boiler and Pressure Vessel Code is the appropriate Code requirement that contains the Charpy-V, RTNDT, and lowest service temperature (RT.,

+100*F) requirements for materials used to fabricate pressure retaining pipes, valves, and pumps in the reactor coolant pressure boundary. We -

recommend that a new reference to Paragraph NS-2332 (1986 Edition of Section lil) be added to the list of documents in proposed TS 5.9.6 to reflect this. NOTE: in the plant specific PTLR, OPPD is using RT., br the pipes, 50*F +120*F + 12*F instrument uncertainty - this is conservative to the  ?!

requirements of NB-2332 and is acceptable.

I 1

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iii. It needs to be emphasized that the staff cannot approve a general reference to the latert approved version of the CE NPSD-683 because there is no requirement to resubmit the report for review and approval if the methodologies in the report get reviced by CE at a later date. The reference to methodologies in Topical Report No. CE NPSD-683 therefore has to be to the specific version of topical report approved by the staff. The NRC staff will also require that the reference to the topical report also contain a referance to the staff's safety evaluation approving the report. The proposed TS 5.9.6 should De amended to reflect these points. The Administrative Control Section for PTLR Programs in Standard CE TS provides a good example of how the new proposed TS 5.9.6 should be worded.

O 2.- Attachment B - Discussion, Justification and No Significant Hazards Consideration

a. Page 1, first and third paragraphs - Reference should be to the specific version of i the topical report, CE NPSD-683, Revision 3.

b. Page 2, second paragraph - OPPD discusses the option of requesting exemptions to use Code Case N-514 or Code Case N-636 at a later date. The reference to Code Case N-636 is incorrect. ASME originally issued the applicable  !

Code Case as Code Case N-626. This Code Case allows a licensee the option of l using the K,e (critical stress intensity value) as the basis for calculating the critical l stress intensity reference values (K,a values) used in the ASME Code Appendix G l P-T limit calculations. Since use of K,e will yield less conservative results than I use of K,4 (the crack arrest stress intensity value), it does not comply with the current version of 10 CFR Part 50, Appendix G; thus an exemption is necessary ,

to use the Code Case. ASME later changed the code case number to N-640.  !

ASME also issued Code Case N-514 to allow a licensee to set the LTOP pressure setpoints at 110% of the maximum allowable pressure from the P-T limit curves; since this is also a relaxation on the requirements of 10 CFR Part 50, Appendix G, an exemption is also needed to use Code Case N-514. The relaxations in Code Case N-514 for LTOP pressure setpcints cannot be requested if a licensee has applied for and been approved to use Code Case N-640 (and visa versa).

OPPD should change the paragraph to correct the code case number from Code Case N-636 to N-640. OPPD should also clarify that an exemption to use Code Case N-514 cannot be applied for if an exemption has been granted to use Code j Case N-640, or visa versa.  !

a 4'

~ c. Page 2, third paragraph - Again, reference should be made to the specific approved version of the topical report, and the paragraph should clarify that a license amendment will be necessary if there is a change to the approved r methodology and OPPD desires to apply the new methodology. -l

'e o .

d. Pages 3 and 4 - Again, references should be made to the specific approved

- version of the topical report, and the paragraph should clarify that a license amendment will be necessary if there is a change to the approved methodology and OPPD desires to apply the new methodology. ,

3. Attachment C - Proposed RCS Pressure - Temperature Limits Report (PTLR)

a. Page 3 - Sec#on 2.2 of the PTLR, " Reactor Vessel Surveillance Program.

Appendix H to 10 CFR Part 50 requires that the programs and withdrawal schedules for reactor vessel matehal surveillance program must meet the requirements of the version of ASTM Standard Procedure E-185 that is current on issue date of the ASME Code to which the reactor vessel was purchased. :The rule gives flexibility to apply the methods and criteria in later versions of the standard procedure, but only through the 1982 version. The rule also states for each capsule withdrawal, the test procedures and reporting requirements for the capsules must meet the requirements of ASTM E-185-82 to the extent practicable

' for the configuration of the specimens in the capsule. i Section 2.2 of the PTLR states that the capsules removed to date have been in accordance with the ASTM E-185 standard procedure in effect at the time of the capsule. The version of E-185 of record for Fort Calhoun is ASTM E-185-66. The current withdrawal schedule for the Fort Calhoun surveillance capsules is provided in Table 4.5-4 of the Fort Calhoun Updated Safety Analysis Report l (USAR). The schedule indicates that three capsules have been pulled from the Fort Calhoun reactor vessel to date (Capsules W-225, W-265 and W-275), with five more capsules scheduled for withdrawal (two at 20 EFPY, one at 21 EFPY, one at 27 EFPY, and one at 32 EFPY). The remaining scheduled capsule withdrawals and tests should done consistent with criteria of ASTM E-185-66 or ASTM E-185-82. It appears that the current schedule in the USAR is consistent with E-185-82. We recommend that OPPD amend the section to reflect that the ,

current program and schedule is consistent with the methods and criteria in ASTM E-185-02, and that the remaining capsules in the program will be pulled in j accordance with the schedule in Table 4.5-4 of the USAR. The section should '

also be amended to reflect that if OPPD opts to change the withdrawal schedule

. In a manner that does not conform to the withdrawal criteria of ASTM E-185-82,

• he changes will be submitted for review and approval (pursuant to the requirements of 10 CFR Part 50, Appendix H, Section Ill.B.3.)

b. Page 5 - Section 2.3.1 states that startup of the first reactor coolant pump with "oold steam generators

• may result in a cooldown which exceeds the 10*F/hr cooldown limit specified in PTLR Figure 4.1, however, per the analysis of CEOO )

Task 1004 Report 0-PENG-ER-012, Revision 01 (e.g., "OOPD Interim Report on 1 ASME Appendix G Evaluation of Step Changes in RCS Temperatures,") this will not result in a cond; tion that exceeds the criteria specified in ASME Section 111 (and XI) or 10 CFR Part 50, Appendix G. The section is confusing and needs i clarification because Figure 4.1 in the PTLR does not provide a 10*F/hr cooldown  !

curve. In regard to the contents of Section 2.3.1 of the PTLR- i i