Text

TSTF Response to NRC Questions on TSTF-564, Safety Limit (Minimum Critical Power Ratio) MCPR
	ML18149A320
Person / Time
Site:	Technical Specifications Task Force
Issue date:	05/29/2018
From:	Gullott D, Miksa J, Sparkman W, Vaughan J, Linda Williams; Technical Specifications Task Force
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TSTF-18-05
	Download: ML18149A320 (66)
	v • d • e

TECHNICAL SPECIFICATIONS TASK FORCE TSTF A JOINT OWNERS GROUP ACTIVITY May 29, 2018 TSTF-18-05 PROJ0753 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

TSTF Response to NRC Questions on TSTF-564, "Safety Limit [Minimum Critical Power Ratio] MCPR" On April 12, 2018, the NRC provided a Request for Information (RAI) regarding TSTF-564, Revision 0, "Safety Limit MCPR."

The TSTF's response to the NRC questions is attached.

The RAI responses discuss changes to TSTF-564. Revision 1 of the traveler is enclosed.

Should you have any questions, please do not hesitate to contact us.

James P. Miksa (PWROG/CE) Lisa L. Williams (BWROG)

David M. Gullott (PWROG/W) Jordan L. Vaughan (PWROG/B&W)

Wesley Sparkman (APOG)

Attachment

Enclosure:

TSTF-564, Revision 1 cc: Michelle Honcharik, Technical Specifications Branch Robert Tjader, Technical Specifications Branch Victor Cusumano, Technical Specifications Branch 11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301-984-4400, Fax: 301-984-7600 Administration by EXCEL Services Corporation

TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR" The NRC comments are repeated below in italics, followed by the TSTF response.

By letter dated August 28, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17240A265), the Technical Specifications Task Force (TSTF) submitted to the U.S. Nuclear Regulatory Commission (NRC) for review Traveler TSTF-564, "Safety Limit MCPR [Minimum Critical Power Ratio]."

1. Section 1 of TSTF-564 states that the proposed change to the MCPR safety limit (SL) is not applicable to plants using AREVA fuel. This is clarified in Section 2.4 of the traveler where it is specifically stated that "the proposed change is not applicable to BWR

[boiling-water reactor] plants using AREVA fuel due to differences in core reload design methodology." It is not acceptable to the NRC staff to categorically exclude fuel from a specific vendor in the manner proposed, because methodologies may be changed and brought into compliance with the intent of the traveler. Instead, please provide the attributes of a core reload design methodology that are necessary for the proposed change to be applicable.

Response

Section 1 is revised to state, "SL MCPR limits are provided for Global Nuclear Fuel and Westinghouse fuel designs. These and other fuel vendors may determine SL MCPR limits for other fuel designs using the described methodology." Conforming changes are made in Section 2.4.

There are only three elements required to apply Equation 1 in Section 3.1 of the TSTF-564 justification, and all three elements are related to the fidelity of the relationship of the critical power (CP) correlation relative to the critical power test data for the specific fuel design. Any fuel vendor must have a fuel-specific CP correlation and must have sufficient relevant CP test data to support it. It is therefore within the ability of any fuel vendor to justify the use of the approach defined in TSTF-564.

The TSTF-564 justification, Section 3.1, Equation 1, utilizes the Experimental Critical Power Ration (ECPR). For each test data point "n", the CP correlation is evaluated at the same conditions as the test to define the Experimental Critical Power Ratio (ECPRn) for that data point as:

( Measured Critical Power )n ECPR n (0.1)

(Critical Power from Correlation)n From all appropriate N of these individual test data points the mean µ and standard deviation of the ECPR distribution is determined. The sample size N determines the value of as explained in Section 3.1 of the TSTF-564 justification. Note that the "i" subscript in Section 3.1 refers to the statistics for a particular BWR fuel product.

This additional explanation of ECPR is added to Section 3.1, providing all the information needed for a fuel vendor to utilize the proposed approach.

Should a licensee wish to use this approach for another fuel vendor, appropriate Bases changes must be developed. Note that the NUREG-1433 and NUREG-1434 BWR Standard Technical Specifications (STS) Chapter 2 Bases currently contain sections 1

TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR" applicable to specific fuel vendors, and not all fuel vendors are included. Including all fuel vendors in the Bases or revising the Bases to provide a more generic description of methodologies would require a significant revision to the Chapter 2 STS Bases that is outside the scope of the traveler.

2. As discussed in paragraph 50.36(c)(1) of Title 10 of the Code of Federal Regulations (10 CFR), plant SLs are "limits upon important process variables that are found necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity." When the SLs are exceeded, the reactor must be shut down to protect the integrity of these barriers and prevent uncontrolled radioactive release.

The pressurized water reactor (PWR) departure from nucleate boiling ratio (DNBR) SL1 has essentially the same basis as the proposed MCPR95/95, in that it represents a 95/95 upper tolerance limit on the correlation used to evaluate the DNBR on the fuel in the core. This translates into a 95 percent probability at a 95 percent confidence level that the fuel will not experience departure from nucleate boiling (DNB) if the SL is not exceeded. The traditional BWR fuel analysis basis is different and requires that 99.9% of the fuel rods in the core are not susceptible to boiling transition.

a. In PWR safety analyses, DNB (and, accordingly, fuel failure) is presumed if the minimum DNBR drops below the DNBR SL. Will exceedance of the MCPR95/95 SL be similarly used as a criterion for boiling transition? If not, please discuss why it is an appropriate SL. Also, please justify why the MCPR operating limits based on the MCPR99.9% SL are compatible with the redefined SL.

b. The NRC staff reviewed current BWR MCPR SLs and found that the SL MCPR95/95 examples provided in Table 1 of TSTF-564 are consistently lower (i.e.,

less conservative) than the current SLMCPR99.9%. Please justify why the basis for these two limits is equivalent or provide a discussion of the unnecessary conservatism included in the MCPR99.9% that is not included in the MCPR95/95.

Response

(a) Yes, the MCPR95/95 Safety Limit will be used as a criterion for boiling transition. As stated in the question, "The pressurized water reactor (PWR) departure from nucleate boiling ratio (DNBR) safety limit has essentially the same basis as the proposed MCPR95/95, in that it represents a 95/95 upper tolerance limit on the correlation used to evaluate the DNBR on the fuel in the core." Boiling transition (BT) nominally occurs when the critical power ratio (CPR) or DNBR reaches a value of unity. The BWR MCPR95/95 and the PWR DNBR95/95 are appropriate Safety Limits because both provide the same 95% probability at 95% confidence based on the correlated test data with high probability and confidence that the lead rod in the core will avoid a calculated boiling crisis in the event of an Anticipated Operational Occurrence (AOO) transient. Both limits serve the same purpose of providing margin to unity 1

Included as SL 2.1.1.1 in the Westinghouse standard technical specifications, NUREG-1431, Volume 1 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A222).

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TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR" where a boiling crisis is calculated to occur by accounting for the correlation bias and uncertainty. The relationship between MCPR99.9% values and the new SLMCPR95/95 is described in part (b) of this response.

(b) Current BWR MCPR SLs (denoted in TSTF-564 as MCPR99.9%) are larger than the SLMCPR95/95 values provided in Table 1 of TSTF-564 because the two are not equivalent. There are two main reasons the SLMCPR95/95 values are lower than the current MCPR99.9% values even though both use the same bias(es) and standard deviation(s) for the CP correlation(s) that apply to the fuel in the core.

(1) In addition to the CP correlation bias and uncertainty, current MCPR99.9% values also include additional uncertainties due to core initial conditions, modeling, and monitoring. These additional uncertainties are combined with the CP values for µ and and the probabilities of individual rods susceptible to boiling transition are summed over all rods in the core to achieve the total number of rods susceptible to boiling transition (NRSBT) equal to 0.1%. The incorporation of plant-specific uncertainties and the summing of individual rod probabilities is what causes the current SLMCPR value to be dependent on both plant- and cycle-specific core design inputs, resulting in frequent license amendments to revise the value.

(2) Neglecting these additional biases and uncertainties and calculating the probability only for the lead rod in the core, for a presumed normal probability density function (PDF), the 99.9% probability (p) at 50% confidence (c) associated with current BWR MCPR SLs corresponds to a value of 3.0902, whereas the SLMCPR95/95 values provided in Table 1 of TSTF-564 utilize a value of approximately 1.7 to 1.8 depending on the sample size N. For a representative example where µ=1, =3.5% and the 95% p/95% c value of

=1.7270 for N=1000, the resulting SLMCPR95/95=1.0604, whereas the same inputs for µ, , and N produce MCPR99.9%=1.1082 because =3.0902 for 99.9% p/50% c. In summary, it is expected that calculated SLMCPR95/95 values are less than currently-calculated MCPR99.9% values.

As stated in Section 1 of TSTF-564, the objective of the traveler is to define a Technical Specification (TS) value that is plant and cycle independent, thereby minimizing the need for frequent license amendment requests to revise the value. The proposed Safety Limit meets that objective and continues to provide protection against postulated fuel damage during AOO transients.

3. The changes proposed in TSTF-564 revise the basis of the SL on critical power ratio such that it "ensures there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling." The critical power ratio is not a measurable quantity in a plant, and must be calculated using a critical power correlation. In effect, the limit is not on the critical power ratio but on the critical power ratio calculated using a particular correlation; the limit is linked to the correlation used to generate it, and is not meaningful without it. Please revise TSTF-564 to include both the value of the revised 3

TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR" MCPR SL and the correlation used to generate it in the technical specification (TS), or provide additional justification that this is not necessary.

Response

The CP correlation(s) used to produce the set of N ECPR values on which the values of µ and are cited in the TSTF-564 proprietary References 1 and 2.

As discussed in the response to Question 14, reference to the proprietary documents describing the CP correlation (which is used to determine both SLMCPR95/95 and MCPR99.9%) will be included in the Core Operating Limits Report (COLR). The TS 3.2.2 operating limit MCPR is currently calculated using the MCPR99.9% value and TSTF-564 does not change that methodology. Therefore, the current COLR references that support determination of the TS 3.2.2 limit also describe the determination of MCPR99.9%.

The TS Bases for the SL are revised to specify the fuel type on which the SL is based.

4. Section 2.4 of TSTF-564 discusses how to determine the SLMCPR95/95 for cores where a mix of applicable fuel designs reside in the core, but does not provide an approach for determining the SLMCPR95/95 for cores that contain a mix of applicable and non-applicable fuel designs. Is TSTF-564 only applicable when the core is completely loaded with applicable assemblies? If not, how is the SLMCPR95/95 determined for cores transitioning to or from applicable fuel designs?

Response

For cores with a mix of fuel products, the corresponding SLMCPR 95/95 is based on the R R largest (i.e., most limiting) of the MCPR 95/95 (i) values for the product lines that are fresh R R or once burnt at the start of the cycle. As indicated in the response to Question 1, the process for defining MCPR95/95 is applicable to any fuel type. A supplier of legacy fuel in a core may be requested to provide the MCPR95/95 value for their fuel or suitable values of database size N, mean µ, and standard deviation , so that the value of MCPR95/95 can be calculated in accordance with Equation 1 of TSTF-564. If the value of N is not supplied, then a value of = 1.8 will be assumed because any viable test database used to develop a CP correlation will have more than 500 data points. The MCPR95/95 for the legacy fuel will be compared to the MCPR95/95 for the new fuel and the most limiting value will be used as the Safety Limit. This comparison may result in a license amendment request if the new SL is more limiting than the existing SL.

5. The concept presented in TSTF-564 for redefining the MCPR SL relies heavily on the experimental critical power ratio (ECPR), which is not mathematically defined in the traveler. Please provide a mathematical definition for ECPR so consistency can be ensured in future licensing actions.

Response

ECPR is defined in Equation 1.1 in the response to Question 1. Section 3.1 of the TSTF-564 justification is revised to add this definition.

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TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR"

6. The NRC staff may find the need to impose safety evaluation penalties on the correlation to ensure appropriate conservatism. As discussed in Information Notice (IN) 2014-01, "Fuel Safety Limit Calculation Inputs Were Inconsistent with NRC-Approved Correlation Limit Values," dated February 21, 2014,2 the NRC staff found instances where the DNBR 95/95 correlation limit was penalized but the penalty was not appropriately fed into plant safety and operating limits. The approach used in the instances found by the NRC staff was similar to that proposed in the traveler, which uses an equation that defines the MCPR95/95 from the ECPR mean, standard deviation, and a statistical factor that is based on the number of data points used to determine the ECPR statistics. How are the penalties imposed on the correlation by the NRC staff (IN 2014-01) reflected in the determination of the MCPR95/95, the MCPR99.9%, and the MCPR operating limit?

Response

The fuel-specific MCPR95/95 depends only on the characteristic of the CP test database for that fuel product. Any penalty associated with a bias in the test database as a whole would be applied to the value of µ and any penalty associated with the characterization of the scatter in the test data as a whole would be applied to . The penalized or adjusted values of µ and would be used to calculate MCPR95/95 for the affected fuel product.

Penalized values of µ and could also affect the MCPR99.9% value, which is used to determine the cycle-specific operating limit MCPR values. If an NRC penalty is related to the process or other inputs to MCPR99.9%, that penalty could also affect the operating limit MCPR values but would not affect the SLMCPR95/95. One example of a bias or uncertainty of this type is associated with the CP correlations ability to represent a particular axial power shape (APS). Another example is monitoring biases or uncertainties that apply only for specific operating domains. Any such penalties and how they are applied would be documented in the methodology documents referenced in the COLR.

7. The TSTF-564 traveler proposes that twice-burnt (or more) fuel is non-limiting with respect to MCPR. Please provide additional justification for this statement, based on theory and/or operating experience, especially for plants operating with short (i.e., 1 year) operating cycles or non-standard operating conditions. Please also discuss whether and how the MCPR95/95 will be updated if a bundle assumed to be non-limiting is found to be limiting during the evaluation of the MCPR99.9% in the development of the MCPR operating limit.

Response

The exclusion of twice-burnt (or greater) fuel (i.e., fuel beginning the third or greater cycle of operation in the reactor core) is based on studies that demonstrate that this fuel will not be limiting with regard to SLMCPR95/95. As indicated in Equation 1 of TSTF-2 ADAMS Accession No. ML13325A966.

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TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR" 564, MCPR95/95(i) depends only on the fuel type (i). Thus, the problem is reduced to identifying the fuel types that potentially could provide the lowest MCPR value.

The current cycle-specific process for calculating SLMCPR considers all fuel in the core and thus requires some form of modeling of the MCPRs for legacy fuel that is in its third cycle (or more) of operation. If this legacy fuel (denoted as "j") is from a different manufacturer than the new fuel, then it is common practice for the fuel transition to use conservative modeling that does not require precise knowledge of the native CP correlation for "j" or its correlation statistics such as µj and j. Such a process has been employed over many cycle-specific evaluations of SLMCPR. It has been proven because even though these later cycle fuel bundles are conservatively biased, they are far from limiting and their probability contribution to the summation that defines NRSBT=0.1% is negligible. (See the part (b) response to Question 2 for additional discussion of NRSBT.)

A mathematical model for estimating relative contributions of bundles to NRSBT is introduced in Section 4.2 of Reference 1 and is further described in the response to Question II-2 starting on page A-18 in Reference 1. This model pertains to any fuel type as described in Eq. (III.2-8) in the RAI response. The difference n between bundle n MCPR (Mn) and the lowest MCPR (M0) in the core is related to the relative probability that bundle n will contribute to NRSBT as depicted in Figure 4-3 of Reference 1.

Figure 4-3 shows that for n 0.08 the relative potential of boiling transition becomes negligible compared to the limiting bundle. How n changes with bundle exposure or bundle residence time depends to some extent on the detailed bundle design, how it is loaded in the core, and how the core is operated.

A representative study considering five BWR cores at beginning, middle, and end of cycle exposures shows that the transient CPR of bundles in their third cycle of operation is 0.14 to 0.54 lower than the CPR of first cycle bundles in these same cores. Overall, the differences were nearer the lower end of this range for 12-month cycles compared to the 18-month and 24-month cycles that were also evaluated; however, even the smallest difference is substantially greater than what is required to safely neglect consideration of any fuel bundles in their third cycle of operation. Therefore, twice-burnt and greater fuel may be neglected when calculating SLMCPR95/95.

Regarding the question related to the operating limit MCPR, MCPR99.9% is used and all fuel bundles are considered. The SLMCPR95/95 is based only on the fuel type and will not be calculated for each fuel cycle. However, the inclusion of cycle-specific factors in the operating limit MCPR ensures that the operating limit MCPR will always be more limiting than the SLMCPR. Also, the TSTF-564 methodology requires that operating limit MCPR (MCPR99.9%) be greater than the SLMCPR95/95. Therefore, the operating limit MCPR evaluation would never identify a need to revise the SLMCPR95/95. If a limiting bundle were to be identified, it would be governed by the more restrictive operating limit MCPR, not the SLMCPR.

8. The TSTF-564 traveler proposes to report the SLMCPR95/95 in the TS to a precision of two digits past the decimal point using standard rounding practices. In the NRC staffs experience, the standard practice for SLs has always been to round up due to the 6

TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR" potential for non-conservatism in the SL introduced by rounding down. Is this the standard rounding practice intended in the traveler? If not, please justify.

Response

Reporting of SLMCPR95/95 will be to two digits past the decimal, rounding up the hundredths digit.

9. The model application requests that licensees discuss the fuels that are or will be in use in the reactor, and discuss which fuel type is limiting with respect to the SLMCPR consistent with the discussion in Section 2.4 of the traveler. The model application then states that SL 2.1.1.2 should be updated with values "consistent with Table 1 of TSTF-564." However, this appears to be inconsistent with the language from the text of Section 2.4 of the traveler, which states, in part, that:

When new fuel types are developed, the fuel vendor will describe to the NRC the derivation of the MCPR95/95 value for that fuel type. This description may be referenced by a licensee requesting a change to SLMCPR95/95.

Such new fuel types would not be referenced in Table 1 of the traveler and thus would constitute a variance from the model application. Please clarify if it is intended that the model application would only be applicable for the fuel types referenced in Table 1.

Response

The model application is revised to state:

The proposed Safety Limit in [SL 2.1.1.2] is [1.07], consistent with [Table 1 of TSTF-564][reference the fuel vendor or NRC documentation of the Safety Limit value for the limiting fuel type].

This change provides an option to reference Table 1 or to reference the fuel vendor documentation or NRC acceptance of a Safety Limit MCPR value for a fuel type not listed Table 1.

10. In the TSTF-564 proposed changes, the second paragraph of the "Background" section of the TS 2.1.1 Bases specifies General Electric and Advanced Nuclear Fuel Corporation (ANF) fuel types. In other sections of the TS Bases, revisions are made to include Westinghouse fuel. Please clarify the discrepancy.

Response

As discussed in the response to Question 1, the NUREG-1433 and NUREG-1434 BWR STS Chapter 2 Bases currently contain sections applicable to specific fuel vendors, and not all fuel vendors are included. Only Bases sections related to the proposed change are revised in TSTF-564. Revising the Chapter 2 Bases to reflect new fuel vendors, renamed fuel vendors, or methodology descriptions not related to the proposed change is outside the scope of the traveler.

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TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR" The TSTF is willing to work with the NRC to update the NUREG-1433 and NUREG-1434 Chapter 2 Bases as a separate, review fee exempt effort.

11. TSTF-564 proposes a language change to the "Background" section of the TS 2.1.1 Bases that defines the SL as ensuring that "at least 99.9% of the fuel rods in the core are not susceptible to boiling transition" rather than "at least 99.9% of the fuel rods in the core do not experience transition boiling." What is the intent of this change, and what is the substance of the difference between these two phrases?

Additionally, an equivalent change to similar language in the "Applicable Safety Analysis" section of the TS 2.1.1 Bases was not made. Please clarify the discrepancy.

Response

The wording "not susceptible to boiling transition" more accurately reflects the statistical process for how MCPR99.9% is calculated, while the phrase "do not experience transition boiling" describes the physical state, not the analytical process. The traveler and Bases, include the Section 2.1.1 Bases, are revised to use the phrase, "not susceptible to boiling transition."

12. In NUREG-1434, Volume 2,3 TS 3.2.2 Bases Reference 2 is to the "current cycle safety analysis." This is appropriate for the discussion in the bases and is applicable regardless of the fuel vendor. In NUREG-1433, Volume 2,4 TS 3.2.2 Bases Reference 2 is to NEDO-24011-P-A, "General Electric Standard Application for Reactor Fuel". Since discussion elsewhere in the TS Bases has been expanded to include Westinghouse fuel, it is not clear that the reference is appropriate. Please clarify.

Response

The STS Bases are used as an example. During conversion to the STS format, a licensee revises the Bases to reflect the plant design and licensing basis. The plant-specific TS Bases are maintained consistent with the plant design and licensing basis using the Technical Specifications Bases Control Program, located in the Administrative Controls section of the TS. Licensees will have revised their TS Bases as needed to reflect changes in fuel supplier and methodology. Therefore, the specific reference in the STS Bases is irrelevant and is only an example of an appropriate reference to be included in the plant-specific Bases.

As discussed in the responses to Questions 1 and 10, only Bases sections related to the proposed change are revised in TSTF-564. Revising the STS Bases for reasons not related to the proposed change is outside the scope of the traveler.

3 Standard Technical Specifications - General Electric Plants (BWR/6): Bases (Revision 4), April 2012 (ADAMS Accession No. ML12104A196).

4 Standard Technical Specifications - General Electric Plants (BWR/4): Bases (Revision 4), April 2012 (ADAMS Accession No. ML12104A193).

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TSTF Response to NRC Questions on TSTF-564, "Safety Limit MCPR"

13. Why is it appropriate to remove "SL" after MCPR in the "Applicability" section of the TS 3.2.2 Bases? With this change, when the "[99.9%]" is not included, the sentence becomes:

Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR is not exceeded even if a limiting transient occurs.

This sentence does not make sense for the MCPR in general, only for a limit placed on the MCPR. Please clarify.

Response

The changes to the sentence are removed and the original sentence is restored, which states, "Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs."

14. Per Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988,5 which established the Core Operating Limits Report (COLR), along with subsequent NRC precedent, the methodologies for determining cycle-specific parameter limits must be included in the list of COLR references contained in TS 5.6.3.b. Given that the MCPR99.9% will now be included in the COLR along with the MCPR operating limit, licensees adopting TSTF-564 must reference appropriate methodologies for the critical power ratio correlation and MCPR99.9% calculation in TS 5.6.3.b. However, this is not discussed in the traveler or model application. Why?

Response

As stated in the traveler justification, the TS 3.2.2 operating limit MCPR is currently calculated using the MCPR99.9% value and TSTF-564 does not change that methodology.

Therefore, the current COLR references that support determination of the TS 3.2.2 limit also describe the determination of MCPR99.9%.

Additional Change in TSTF-564, Revision 1 The NRC review of Westinghouse Optima3 fuel has not been completed. Therefore, references to Optima3 have been removed. Plants desiring to use Optima3 fuel may use the process for new fuel types described in Section 2.4 of the TSTF-564 justification.

Reference:

1. NEDE-32601-P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," August 1999.

5 ADAMS Accession No. ML031130447.

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Enclosure TSTF-564, Revision 1, "Safety Limit MCPR"

BWROG-133, Rev. 0 TSTF-564, Rev. 1 Technical Specifications Task Force Improved Standard Technical Specifications Change Traveler Safety Limit MCPR NUREGs Affected: 1430 1431 1432 1433 1434 2194 Classification: 1) Technical Change Recommended for CLIIP?: Yes Correction or Improvement: Improvement NRC Fee Status: Not Exempt Changes Marked on ISTS Rev 4.0 See attached justification.

Revision History OG Revision 0 Revision Status: Closed Revision Proposed by: BWROG Revision

Description:

Original Issue Owners Group Review Information Date Originated by OG: 30-May-17 Owners Group Comments Reviewed by the Owners Groups in October 2016.

A presubmittal teleconfernce was held on March 20, 2017. The traveler was revised to address NRC comments:

Owners Group Resolution: Approved Date: 18-Jul-17 TSTF Review Information TSTF Received Date: 02-Aug-17 Date Distributed for Review 02-Aug-17 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 28-Aug-17 NRC Review Information NRC Received Date: 28-Aug-17 NRC Comments:

Revised to reflect RAI responses.

Final Resolution: NRC Requests Changes: TSTF Will Revise 29-May-18 Copyright(C) 2018, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-133, Rev. 0 TSTF-564, Rev. 1 TSTF Revision 1 Revision Status: Active Revision Proposed by: BWROG Revision

Description:

Revised to reflect RAI response:

The justification and Bases markup were revised and expanded to discuss application of the methodology to other fuel vendors.

The changes to the TS 3.2.2 Applicability Bases were removed.

Discussion of boiling transition was made more consistent in the justification and Bases.

Reference to Westinghouse Optima3 fuel was removed.

The SLMCPR value will be rounded up.

The TS Bases states the fuel type on which the SLMCPR value is based.

Owners Group Review Information Date Originated by OG: 10-Apr-18 Owners Group Comments (No Comments)

Owners Group Resolution: Approved Date: 02-May-18 TSTF Review Information TSTF Received Date: 10-Apr-18 Date Distributed for Review 10-Apr-18 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 02-May-18 NRC Review Information NRC Received Date: 29-May-18 Affected Technical Specifications Bkgnd 2.1.1 Bases Reactor Core SLs S/A 2.1.1 Bases Reactor Core SLs SL 2.1.1.2 Safety Limit MCPR Bkgnd 3.2.2 Bases Minimum Critical Power Ratio (MCPR)

S/A 3.2.2 Bases Minimum Critical Power Ratio (MCPR)

LCO 3.2.2 Bases Minimum Critical Power Ratio (MCPR) 29-May-18 Copyright(C) 2018, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

BWROG-133, Rev. 0 TSTF-564, Rev. 1 Appl. 3.2.2 Bases Minimum Critical Power Ratio (MCPR) 5.6.3 Core Operating Limits Report 29-May-18 Copyright(C) 2018, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-564, Rev. 1

1.

SUMMARY

DESCRIPTION The Technical Specification (TS) Safety Limit (SL) value and method of calculation for the Minimum Critical Power Ratio (MCPR) limit, SL 2.1.1.2, is revised for Boiling Water Reactor (BWR) plants. SL MCPR limits are provided for Global Nuclear Fuel and Westinghouse fuel designs. These and other fuel vendors may determine SL MCPR limits for other fuel designs using the described methodology. The revised calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95%

confidence level that no rods will be susceptible to transition boiling. The revised MCPR SL is consistent with the regulatory requirements while being cycle-independent, thereby minimizing the need for TS license amendment requests to revise this value for each operating cycle.

2. DETAILED DESCRIPTION 2.1. Current Design and Licensing Basis MCPR is defined in Section 1.1 of the BWR TS as:

The MCPR shall be the smallest critical power ratio (CPR) that exists in the core [for each class of fuel]. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations are developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Although fuel damage does not necessarily occur if a fuel rod experiences transition boiling, the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient metric for ensuring that fuel failures due to inadequate cooling do not occur.

The TS contain two limits on MCPR: a safety limit (herein referred to as the SLMCPR) and a Limiting Condition for Operation (LCO) operating limit (herein referred to as the OLMCPR).

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, paragraph 50.36(c)(1) defines "safety limits" as limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. In the case of the MCPR safety limit, the physical barrier being protected is the fuel rod cladding. The current SLMCPR is calculated as the point at which 99.9% of the fuel rods are not susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences, herein referred to as MCPR99.9%. The MCPR99.9% limit is calculated for each BWR on a cycle-by-cycle basis using approved methodologies.

Page 1

TSTF-564, Rev. 1 An LCO is defined in 10 CFR 50.36(c)(2) as the lowest functional capability or performance level of equipment required for safe operation of the facility. The OLMCPR LCO is required to be met to ensure that no fuel damage results during anticipated operational occurrences (AOOs).

To ensure that the measured MCPR does not exceed the SLMCPR during any AOO that occurs with moderate frequency, transients are analyzed to determine the largest reduction in critical power ratio. The limiting transient yields the largest change in CPR (CPR) during the event.

The largest CPR is combined with the MCPR99.9% value to determine the OLMCPR LCO limit.

Together, SLMCPR and OLMCPR ensure that no fuel damage occurs due to transition boiling during normal operation or AOOs.

2.2. Current Technical Specifications Requirements NUREG-1433 and NUREG-14341, Safety Limit 2.1.1.2 states:

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be [1.07] for two recirculation loop operation or [1.08] for single recirculation loop operation.

The values in brackets, [1.07] and [1.08], are plant-specific limits. The reactor steam dome pressure and core flow values are also plant-specific and differences do not affect the applicability of the proposed change.

NUREG-1433 and NUREG-1434, LCO 3.2.2, "Minimum Critical Power Ratio (MCPR)," states:

All MCPRs shall be greater than or equal to the MCPR operating limits specified in the

[Core Operating Limits Report] COLR.

NUREG-1433 and NUREG-1434, LCO 3.2.2 is applicable when thermal power is 25% of rated thermal power. Plant-specific TS may have a different Applicability, which does not affect the justification for the proposed change.

1 NUREG-1433 is based on the BWR/4 plant design, but is also applicable of the BWR/2, BWR/3, and, for some requirements, to the BWR/5 plant designs. NUREG-1434 is based on the BWR/6 plant design, and is applicable, for some requirements, to the BWR/5 plant design.

Page 2

TSTF-564, Rev. 1 NUREG-1433 and NUREG-1434, TS 5.6.3, "Core Operating Limits Report," states:

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

[The individual specifications that address core operating limits must be referenced here.]

2.3. Reason for the Proposed Change The current SLMCPR (i.e., MCPR99.9%) is affected by the cycle-specific design, such as core power distribution, fuel type, and operating power-flow domain. These factors generally vary enough from cycle-to-cycle that changes to the SLMCPR TS values are common. The subsequent cycle core design is dependent on the core burnup of the previous cycle. As a result, the core design for the subsequent cycle is typically finalized late in the previous fuel cycle.

Consequently, license amendment requests to modify the SLMCPR typically request an accelerated NRC review (i.e., less than the typical period of one year) to support the scheduled start of the subsequent fuel cycle. A review of the NRC Agency-wide Documents Access and Management System (ADAMS) identified a number of approved license amendments to revise the SLMCPR in 2015 and 2016.

The Babcock & Wilcox, Westinghouse, Combustion Engineering, and Advanced Passive 1000 (AP1000) Standard Technical Specifications (NUREG-1430, NUREG-1431, NUREG-1432, and NUREG-2194) safety limits on fuel cladding are based on a Departure from Nucleate Boiling Ratio (DNBR) limit. The PWR DNBR limits are roughly analogous to the BWR SLMCPR, in that both protect fuel cladding integrity from inadequate cooling. The PWR DNBR safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur, vice the BWR SLMCPR that is based on ensuring that 99.9% of the fuel rods will not be susceptible to boiling transition. Either approach is statistically valid, but this difference results in a PWR safety limit that is only dependent on the fuel type(s) in the reactor and the corresponding DNBR correlations. The PWR DNBR Safety Limits are not cycle dependent and are typically only revised when the type of fuel changes.

2.4. Description of the Proposed Change The proposed change revises the standard TS in NUREG-1433 and NUREG-1434 for all BWR plants. SL MCPR limits are provided for Global Nuclear Fuel and Westinghouse fuel designs.

These and other fuel vendors may determine SL MCPR limits for other fuel designs using the described methodology.

The proposed change substantially reduces the need for cycle-specific changes to the SLMCPR and eliminates the need for accelerated NRC review of those changes.

Page 3

TSTF-564, Rev. 1 The NUREG-1433 and NUREG-1434 Safety Limit 2.1.1.2 is revised to state:

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be [1.07] [for two recirculation loop operation or [1.08] for single recirculation loop operation].

The phrase "for two recirculation loop operation or [1.08] for single recirculation loop operation" are shown in brackets to retain compatibility for BWR plants that do not adopt the proposed change. The proposed SLMCPR methodology is not dependent on the number of recirculation loops in operation, so the distinction between a single loop and two loop operation is not needed.

Plants adopting the proposed change will revise their plant-specific SL to state:

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10% rated core flow:

MCPR shall be [1.07].

The bracketed limit "[1.07]" will be replaced with a revised SLMCPR that ensures there is a 95%

probability at a 95% confidence level that no rods will be susceptible to transition boiling, and is referred to as SLMCPR95/95. The reactor steam dome pressure and core flow values are also plant-specific. Differences between the Standard Technical Specifications values and the plant-specific values do not affect the applicability of the proposed change.

The single SLMCPR95/95 is based on the fuel type in the reactor core.

Table 1: Proposed MCPR95/95 Values by Vendor and Fuel Bundle Type Vendor Fuel Type Proposed MCPR95/95 Global GE14 1.05 Nuclear Fuel Global GNF2 1.07 Nuclear Fuel Global GNF3 1.07 Nuclear Fuel Westinghouse Optima2 1.06 The derivation of these values is described in proprietary letters to the NRC from Global Nuclear Fuel and Westinghouse (References 1 and 2). Should these or different vendors desire to use this Page 4

TSTF-564, Rev. 1 approach for existing or new fuel types, the fuel vendor will describe to the NRC the derivation of the MCPR95/95 value for that fuel type. This description may be referenced by a licensee requesting a change to SLMCPR95/95. The TS Bases state the fuel type on which the SL is based.

For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR95/95 values for the fuel products that are fresh or once-burnt at the start of the cycle.

LCO 3.2.2 is not affected by the proposed change. However, licensees adopting the proposed change will include the MCPR99.9% value (i.e., the value equivalent to the current, cycle-dependent SLMCPR) in the COLR values for LCO 3.2.2.

TS 5.6.3, "Core Operating Limits Report," paragraph a, is revised to require the MCPR99.9%

value to be in the cycle-specific COLR:

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following (addition shown in italics):

[The individual specifications that address core operating limits must be referenced here.

The MCPR99.9% value used to calculate the LCO 3.2.2, "MCPR," limit shall be specified in the COLR.]

The proposed change is supported by changes to the TS Bases. The SL 2.1.1.2 Bases and TS 3.2.2 Bases are revised to reflect the proposed limits for Global Nuclear Fuel and Westinghouse fuel. In sections of the Bases applicable to all fuel types, the existing text and the proposed text are both presented in brackets, signifying that the licensee should choose the applicable description. A reviewer's note is added to the Bases to explain these options. The regulation at 10 CFR 50.36, states, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." A licensee may make changes to the TS Bases without prior NRC staff review and approval in accordance with the Technical Specifications Bases Control Program. The proposed TS Bases changes are consistent with the proposed TS changes and provide the purpose for each requirement in the specification consistent with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 1993 (58 FR 39132). Therefore, the Bases changes are provided for information and approval of the Bases is not requested.

A model application is included in the proposed change as Enclosure 1. The model may be used by licensees desiring to adopt the traveler following NRC approval.

3. TECHNICAL EVALUATION The proposed change revises the TS limit for the SLMCPR for the applicable plants and places the existing SLMCPR value (i.e., MCPR99.9%) in the COLR. The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition Page 5

TSTF-564, Rev. 1 boiling. The revised SLMCPR, referred to as SLMCPR95/95, is based only on the CPR correlation uncertainty determined for the Global Nuclear Fuel or Westinghouse fuel type. Plant and cycle-specific uncertainties are not included in the SLMCPR95/95. These uncertainties are currently and will continue to be included in the OLMCPR LCO. Reactor coolant flow is one of the uncertainties removed from the SLMCPR calculation and retained in the OLMCPR.

Therefore, the SLMCPR for dual recirculation loop operation and single recirculation loop operation are replaced with a single SLMCPR95/95.

The LCO 3.2.2 limits (i.e., the OLMCPR values) are not changed and will be based on the existing SLMCPR, referred to as MCPR99.9%. The OLMCPR will continue to be determined based on the transient CPR components and the cycle-specific MCPR99.9% value that will be included in the COLR. Therefore, the margin to boiling transition remains unchanged.

3.1. Statistical Treatment of MCPR95/95 For each Global Nuclear Fuel and Westinghouse BWR fuel product (designated i), the MCPR95/95(i) is calculated using that products experimentally determined critical power statistics as follows:

MCPR95/95(i) = µi + i*i (Eq. 1)

Where,

µi is the mean Experimental Critical Power Ratio (ECPR),

i is the standard deviation of the ECPRs, and i is a statistical parameter chosen to provide 95% probability at 95% confidence (95/95) for the one-sided upper tolerance limit that depends on the number of samples (Ni) in the critical power database.

i is a fuel product line, such as GE14, GNF2, GNF3, and OPTIMA2.

For each test data point "n", the Critical Power (CP) correlation is evaluated at the same conditions as the test to define the Experimental Critical Power Ratio (ECPRn) for that data point as:

( Measured Critical Power )n ECPR n (Eq. 1.1)

(Critical Power from Correlation)n From all appropriate N of these individual test data points the mean and standard deviation of the ECPR distribution is determined. The sample size N determines the value of .

The statistical parameter, i, is calculated using formulas attributed to Mary Gibbons Natrella (1963) as recommended by the National Institute of Standards and Technology (NIST) in their Page 6

TSTF-564, Rev. 1 Engineering Statistics Handbook (Reference 3). For a 95/95 probability/confidence level, the i values are shown in the table below as a function of database size (Ni).

Table 2: Statistical Parameter, i, at (95/95) for the One-Sided Upper Tolerance Limit Database Size, Ni i 500 1.7625 750 1.7401 1000 1.7270 1250 1.7181 1500 1.7115 2000 1.7024 Assuming a typical critical power database of 1000 data points with no bias (i.e., µi = 1.0), the following table illustrates representative MCPR95/95(i) values as a function of the database standard deviation.

Table 3: Representative MCPR95/95 Values for Ni=1000 Standard MCPR95/95 Deviation, i (%)

2.0 1.03 2.5 1.04 3.0 1.05 3.5 1.06 4.0 1.07 4.5 1.08 5.0 1.09 For cores loaded with a single fuel product, the SLMCPR95/95 is the MCPR95/95(i) value for that particular product line.

Page 7

TSTF-564, Rev. 1 For cores with a mix of fuel products, the corresponding SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR95/95(i) values for the product lines that are fresh or once-burnt at the start of the cycle. The MCPR95/95(i) values for product lines that are twice-burnt or more at the start of the cycle may be ignored, as these higher exposure bundles operate with considerable MCPR margin relative to the more limiting fresh and once-burnt bundles.

The SLMCPR95/95 will be reported in the TS to two digits past the decimal, rounding up. The SLMCPR95/95 also serves as the minimum value for the cycle-specific MCPR99.9%.

The revised method for calculation of the SL for Global Nuclear Fuel and Westinghouse fuel will continue to meet the regulatory definition of a safety limit and to reasonably protect the integrity of the fuel rod cladding against the uncontrolled release of radioactivity. The proposed change is also consistent with equivalent safety limits for other plant designs.

3.2. Cycle-Specific OLMCPR The current MCPR99.9% statistical limit calculation will continue to be performed using the approved methodology (e.g., References 4 through 7 or the plant-specific equivalents) and will be reported in the COLR. The OLMCPR limit in LCO 3.2.2 will continue to be determined based on the transient CPR component and the cycle-specific MCPR99.9% value. No changes to the method of determining the OLMCPR (i.e., the LCO 3.2.2 limit) are proposed, and the LCO limits and the MCPR99.9% value will be reported in the COLR.

4. REGULATORY EVALUATION Title 10 of the Code of Federal Regulations (10 CFR), Part 50, paragraph 50.36(c)(1)(i)(A) states:

Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.

The purpose of the MCPR safety limit (SLMCPR) is to protect the physical barrier of the fuel cladding against the uncontrolled release of radioactivity. The SLMCPR is set such that no significant fuel damage is calculated to occur if the limit is met. Although it is recognized that the onset of transition boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. Therefore, the proposed change to the SLMCPR will continue to protect the fuel cladding physical barrier from uncontrolled release of radioactivity.

10 CFR 50, Appendix A, General Design Criterion 10 states that specified acceptable fuel design limits will not be exceeded during steady state operation, normal operational transients, and Anticipated Operational Occurrences (AOOs). Most plants have a plant-specific design criterion Page 8

TSTF-564, Rev. 1 similar to GDC 10. This design criterion will continue to be met. The OLMCPR, which is not affected by the proposed change, is established to ensure that no fuel damage results during normal operation, normal operational transients, and AOOs.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

5. REFERENCES

1. Letter from Brian R. Moore, Global Nuclear Fuel, to U.S. NRC, "Information Supporting TSTF-564 Safety Limit Minimum Critical Power Ratio," June 16, 2017, ADAMS Accession No. ML17167A108.

2. Letter from James A. Gresham, Westinghouse Electric Company, to U.S. NRC, "Submittal of 'Calculation for Technical Specification SLM CPR Values Applying to Westinghouse Fuel in Support of TSTF-564'," May 16, 2017, ADAMS Accession No. ML17142A319.

3. NIST/SEMATECH e-Handbook of Statistical Methods, http://www.itl.nist.gov/div898/handbook/, April 2012.

4. GE Nuclear Energy, "General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NEDO-10958-A, January 1977, ADAMS Accession No. ML102290144.

5. GE Nuclear Energy, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," NEDC-32694P-A, August 1999, ADAMS Accession No. ML003740166.

6. GE Nuclear Energy, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," NEDC-32601-P-A, August 1999, ADAMS Accession No. ML003740166.

7. ABB Combustion Engineering Nuclear Operations, "Reference Safety Report for Boiling Water Reactor Reload Fuel," CENPD-300-P-A, July 1996. ADAMS Accession No. ML110260388.

Page 9

TSTF-564, Rev. 1 Enclosure 1 Model Application

TSTF-564, Rev. 1

[DATE] 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 DOCKET NO.PLANT NAME 50-[xxx]

SUBJECT:

Application to Revise Technical Specifications to Adopt TSTF-564, "Safety Limit MCPR" Pursuant to 10 CFR 50.90, [LICENSEE] is submitting a request for an amendment to the Technical Specifications (TS) for [PLANT NAME, UNIT NOS.].

[LICENSEE] requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the [PLANT NAME, UNIT NOS] Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked to show the proposed changes for information only.

No regulatory commitments are made in this submittal.

Approval of the proposed amendment is requested by [date]. Once approved, the amendment shall be implemented within [ ] days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official.

[In accordance with 10 CFR 50.30(b), a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by attaching a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct.

Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.]

If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER].

Sincerely, Page 1

TSTF-564, Rev. 1

[Name, Title]

Attachments: 1. Description and Assessment

2. Proposed Technical Specification Changes (Mark-Up)

3. Revised Technical Specification Pages

4. Proposed Technical Specification Bases Changes (Mark-Up) for Information Only

{Attachments 2, 3, and 4 are not included in the model application and are to be provided by the licensee.}

cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Page 2

TSTF-564, Rev. 1 ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

[LICENSEE] requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the [PLANT NAME, UNIT NOS] Technical Specifications (TS). The proposed amendment revises the Technical Specification (TS) safety limit (SL) on minimum critical power ratio (MCPR) to reduce the need for cycle-specific changes to the value while still meeting the regulatory requirement for an SL.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation

[LICENSEE] has reviewed the safety evaluation for TSTF-564 provided to the Technical Specifications Task Force in a letter dated [DATE]. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-564. [As described herein,]

[LICENSEE] has concluded that the justifications presented in TSTF-564 and the safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the [PLANT] TS.

The [PLANT], Unit [Unit Numbers], reactor [is currently][will be] fueled with [TYPE] fuel bundles [describe multiple types of fuel bundles and which type limits the SLMCPR consistent with discussion in the traveler]. The proposed Safety Limit in [SL 2.1.1.2] is [1.07], consistent with [Table 1 of TSTF-564][ reference the fuel vendor or NRC documentation of the Safety Limit value for the limiting fuel type].

The MCPR value calculated as the point at which 99.9% of the fuel rods would not be susceptible to boiling transition (i.e., reduced heat transfer) during normal operation and anticipated operational occurrences is referred to as MCPR99.9%. Technical Specification 5.6.3, "Core Operating Limits Report (COLR)," is revised to require the MCPR99.9% value to be included in the cycle-specific COLR.

2.2 Variations

((LICENSEE] is not proposing any variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation dated [DATE].] ((LICENSEE] is proposing the following variations from the TS changes described in TSTF-564 or the applicable parts of the NRC staffs safety evaluation: describe the variations]

[The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical Specifications on which TSTF-564 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-564 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-564 to the [PLANT]

TS.]

[The [PLANT] TS contain requirements that differ from the Standard Technical Specifications on which TSTF-564 was based, such as reactor steam dome pressure or core flow in SL 2.1.1.2, Page 3

TSTF-564, Rev. 1 or Applicability in TS 3.2.2, but these differences do not affect the applicability of the TSTF-564 justification. [For differences other than reactor steam dome pressure, core flow, or applicability, describe the differences and why TSTF-564 is still applicable.]

[The traveler and Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). [PLANT] was not licensed to the 10 CFR 50, Appendix A, GDC. The [PLANT] equivalents of the referenced GDC are [reference including UFSAR location, if applicable]. [Discuss the equivalence of the referenced plant-specific requirements to the Appendix A GDC as related to the proposed change.] This difference does not alter the conclusion that the proposed change is applicable to

[PLANT].]

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis

[LICENSEE] requests adoption of TSTF-564, "Safety Limit MCPR," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the [PLANT NAME, UNIT NOS] Technical Specifications (TS). The proposed change revises the Technical Specifications (TS) safety limit on minimum critical power ratio (SLMCPR). The revised limit calculation method is based on using the Critical Power Ratio (CPR) data statistics and is revised from ensuring that 99.9% of the rods would not be susceptible to boiling transition to ensuring that there is a 95% probability at a 95% confidence level that no rods will be susceptible to transition boiling. A single SLMCPR value will be used instead of two values applicable when one or two recirculation loops are in operation. TS 5.6.3, "Core Operating Limits Report (COLR)," is revised to require the current SLMCPR value to be included in the COLR.

[LICENSEE] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the Core Operating Limits Report (COLR). The SLMCPR is not an initiator of any accident previously evaluated. The revised safety limit values continue to ensure for all accidents previously evaluated that the fuel cladding will be protected from failure due to transition boiling. The proposed change does not affect plant operation or any procedural or administrative controls on plant operation that affect the functions of preventing or mitigating any accidents previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Page 4

TSTF-564, Rev. 1

2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. The proposed change will not affect the design function or operation of any structures, systems or components (SSCs). No new equipment will be installed. As a result, the proposed change will not create any credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed amendment revises the TS SLMCPR and the list of core operating limits to be included in the COLR. This will result in a change to a safety limit, but will not result in a significant reduction in the margin of safety provided by the safety limit. As discussed in the application, changing the SLMCPR methodology to one based on a 95%

probability with 95% confidence that no fuel rods experience transition boiling during an anticipated transient instead of the current limit based on ensuring that 99.9% of the fuel rods are not susceptible to boiling transition does not have a significant effect on plant response to any analyzed accident. The SLMCPR and the TS Limiting Condition for Operation (LCO) on MCPR continue to provide the same level of assurance as the current limits and do not reduce a margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, [LICENSEE] concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Page 5

TSTF-564, Rev. 1 4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Page 6

TSTF-564, Rev. 1 Enclosure 2 Technical Specifications Proposed Changes

TSTF-564, Rev. 1 SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall be 25% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%

rated core flow:

MCPR shall be [1.07] [for two recirculation loop operation or [1.08]

for single recirculation loop operation.]

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1325 psig.

2.2 SL VIOLATIONS With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

General Electric BWR/4 STS 2.0-1 Rev. 4.0

TSTF-564, Rev. 1 MCPR 3.2.2 No Changes. Included for Reference 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY: THERMAL POWER 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits. limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to < 25% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 25% RTP AND

[ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/4 STS 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 1 MCPR 3.2.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.2.2.2 Determine the MCPR limits. Once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each completion of SR 3.1.4.1 AND Once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each completion of SR 3.1.4.2 AND Once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each completion of SR 3.1.4.4 General Electric BWR/4 STS 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 1 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

[ The individual specifications that address core operating limits must be referenced here. The MCPR99.9% value used to calculate the LCO 3.2.2, "MCPR," limit shall be specified in the COLR.]

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ]

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]

General Electric BWR/4 STS 5.6-2 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for [both General Electric Company (GE) and Advanced Nuclear Fuel Corporation (ANF) fuel]. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

REVIEWER'S NOTE ------------------------------

In the Background and Applicable Safety Analysis sections, select the SLMCPR95/95 discussion or the 99.9% of the fuel rods discussion as the applicable SL 2.1.1.2 basis.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. [This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95/95, which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.] [The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core are not susceptible to boiling transitiondo not experience transition boiling].

General Electric BWR/4 STS B 2.1.1-1 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

General Electric BWR/4 STS B 2.1.1-2 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. [The Tech Spec SL is set generically on a fuel ANALYSES product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95] [The reactor core SLs are established to preclude violation of the fuel design criterion that a MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be susceptible to boiling transitionexpected to experience the onset of transition boiling.]

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1.1a Fuel Cladding Integrity [General Electric Company (GE) Fuel]

GE critical power correlations are applicable for all critical power calculations at pressures 785 psig and core flows 10% of rated flow.

For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER > 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.

2.1.1.1b Fuel Cladding Integrity [Advanced Nuclear Fuel Corporation (ANF) Fuel]

The use of the XN-3 correlation is valid for critical power calculations at pressures > 580 psig and bundle mass fluxes > 0.25 x 106 lb/hr-ft2 (Ref. 3). For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For the General Electric BWR/4 STS B 2.1.1-3 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 ANF 9x9 fuel design, the minimum bundle flow is > 30 x 103 lb/hr. For the ANF 8x8 fuel design, the minimum bundle flow is > 28 x 103 lb/hr. For all designs, the coolant minimum bundle flow and maximum flow area are General Electric BWR/4 STS B 2.1.1-4 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) such that the mass flux is always > 0.25 x 106 lb/hr-ft2. Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 106 lb/hr-ft2 is approximately 3.35 MWt.

At 25% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of > 3.0, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 25% RTP for reactor pressures < 785 psig is conservative.

2.1.1.2a MCPR [GE and Westinghouse Fuel]

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. [The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.]

Reviewer's Note --------------------------------

The MCPR95/95 Values by Vendor and Fuel Product Type:

Vendor Fuel Type MCPR95/95 Global GE14 1.05 Nuclear Fuel Global GNF2 1.07 Nuclear Fuel Global GNF3 1.07 Nuclear Fuel Westinghouse Optima2 1.06

[The SL is based on [Fuel Type] fuel. For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle.]

[However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which General Electric BWR/4 STS B 2.1.1-5 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.]

2.1.1.2b MCPR [ANF Fuel]

The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed statistical General Electric BWR/4 STS B 2.1.1-6 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the XN-3 critical power correlation. Reference 3 describes the methodology used in determining the MCPR SL.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the XN-3 correlation, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that there would be no transition boiling in the core during sustained operation at the MCPR SL.

If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

General Electric BWR/4 STS B 2.1.1-7 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. NEDE-24011-P-A (latest approved revision).

3. XN-NF524(A), Revision 1, November 1983.

4. 10 CFR 100.

General Electric BWR/4 STS B 2.1.1-8 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2. The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs), and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated.

Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE ----------------------------------- REVIEWER'S NOTE -------------------------------

SAFETY Incorporate the MCPR95/95 discussion if applicable.

ANALYSES ---------------------------------------------------------------------------------------------.--

The analytical methods and assumptions used in evaluating the AOOs to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 6, 7, and 8. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (CPR). When the largest CPR is combined with added to the [SL] MCPR[99.9%] SL, the required operating limit MCPR is obtained.

[MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical General Electric BWR/4 STS B 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9%

statistical analysis.]

The MCPR operating limits are derived from [the MCPR99.9% value and]

the transient analysis, and are dependent on the operating core flow and power state (MCPRf and MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 6, 7, and 8). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 9) to analyze slow flow runout transients. The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

General Electric BWR/4 STS B 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 BASES APPLICABLE SAFETY ANALYSES (continued)

Power dependent MCPR limits (MCPRp) are determined by approved transient analysis modelsmainly by the one dimensional transient code (Ref. 10). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The MCPR operating limits specified in the COLR [(MCPR99.9% value, MCPRf values, and MCPRp values)] are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPRf and MCPRp limits[, which are based on the MCPR99.9% limit specified in the COLR.]

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels

< 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the General Electric BWR/4 STS B 3.2.2-3 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

General Electric BWR/4 STS B 3.2.2-4 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 BASES ACTIONS (continued)

B.1 If the MCPR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is 25% RTP and periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

[ The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.

SR 3.2.2.2 determines the value of , which is a measure of the actual scram speed distribution compared with the assumed distribution. The MCPR operating limit is then determined based on an interpolation General Electric BWR/4 STS B 3.2.2-5 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued) between the applicable limits for Option A (scram times of LCO 3.1.4, "Control Rod Scram Times") and Option B (realistic scram times) analyses. The parameter must be determined once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is acceptable due to the relatively minor changes in expected during the fuel cycle.

REFERENCES 1. NUREG-0562, June 1979.

2. NEDO-24011-P-A, "General Electric Standard Application for Reactor Fuel" (latest approved version).

3. FSAR, Chapter [4].

4. FSAR, Chapter [6].

5. FSAR, Chapter [15].

6. [Plant specific single loop operation].

7. [Plant specific load line limit analysis].

8. [Plant specific Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvements (ARTS) Program].

9. NEDO-30130-A, "Steady State Nuclear Methods," May 1985.

10. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.

General Electric BWR/4 STS B 3.2.2-6 Rev. 4.0

TSTF-564, Rev. 1 SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall be 25% RTP.

2.1.1.2 With the reactor steam dome pressure 785 psig and core flow 10%

rated core flow:

MCPR shall be [1.07] [for two recirculation loop operation or [1.08]

for single recirculation loop operation.]

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 1325 psig.

2.2 SL VIOLATIONS With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

General Electric BWR/6 STS 2.0-1 Rev. 4.0

TSTF-564, Rev. 1 MCPR 3.2.2 No Changes. Included for Reference 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.

APPLICABILITY: THERMAL POWER 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limits. limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to < 25% RTP.

Time not met.

General Electric BWR/6 STS 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 1 MCPR 3.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified in the COLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 25% RTP AND

[ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/6 STS 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 1 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

[ The individual specifications that address core operating limits must be referenced here. The MCPR99.9% value used to calculate the LCO 3.2.2, "MCPR," limit shall be specified in the COLR.]

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

REVIEWERS NOTE----------------------------------------

Licensees that have received prior NRC approval to relocate Topical Report revision numbers and dates to licensee control need only list the number and title of the Topical Report, and the COLR will contain the complete identification for each of the Technical Specification referenced Topical Reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). See NRC ADAMS Accession No: ML110660285 for details.

[ Identify the Topical Report(s) by number, title, date, and NRC staff approval document or identify the staff Safety Evaluation Report for a plant specific methodology by NRC letter and date. ]

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

[ The individual specifications that address RCS pressure and temperature limits must be referenced here. ]

General Electric BWR/6 STS 5.6-2 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for [both General Electric Company (GE) and Advanced Nuclear Fuel Corporation (ANF) fuel]. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

REVIEWER'S NOTE ------------------------------

In the Background and Applicable Safety Analysis sections, select the SLMCPR95/95 discussion or the 99.9% of the fuel rods discussion as the applicable SL 2.1.1.2 basis.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. [This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95/95, which corresponds to a 95% probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.] [The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core are not susceptible to boiling transitiondo not experience transition boiling.]

General Electric BWR/6 STS B 2.1.1-1 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

General Electric BWR/6 STS B 2.1.1-2 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. [The Tech Spec SL is set generically on a fuel ANALYSES product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95.] [The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core are not susceptible to boiling transitionwould not be expected to experience the onset of transition boiling.]

The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.

2.1.1.1a Fuel Cladding Integrity [General Electric Company (GE) Fuel]

GE critical power correlations are applicable for all critical power calculations at pressures 785 psig and core flows 10% of rated flow.

For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER > 50% RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.

2.1.1.1b Fuel Cladding Integrity [Advanced Nuclear Fuel Corporation (ANF) Fuel]

The use of the XN-3 correlation is valid for critical power calculations at pressures > 580 psig and bundle mass fluxes > 0.25 x 106 lb/hr-ft2 (Ref. 3). For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a General Electric BWR/6 STS B 2.1.1-3 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 relatively high power and potentially can approach a critical heat flux condition. For the ANF 9x9 fuel design, the minimum bundle flow General Electric BWR/6 STS B 2.1.1-4 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) is > 30 x 103 lb/hr. For the ANF 8x8 fuel design, the minimum bundle flow is > 28 x 103 lb/hr. For all designs, the coolant minimum bundle flow and maximum flow area are such that the mass flux is always

> 0.25 x 106 lb/hr-ft2. Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 106 lb/hr-ft2 is approximately 3.35 MWt. At 25% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of > 3.0, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 25% RTP for reactor pressures < 785 psig is conservative.

2.1.1.2a MCPR [GE and Westinghouse Fuel]

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. [The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.]

Reviewer's Note --------------------------------

The MCPR95/95 Values by Vendor and Fuel Product Type:

Vendor Fuel Type MCPR95/95 Global GE14 1.05 Nuclear Fuel Global GNF2 1.07 Nuclear Fuel Global GNF3 1.07 Nuclear Fuel Westinghouse Optima2 1.06

[The SL is based on [Fuel Type] fuel. For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle.]

General Electric BWR/6 STS B 2.1.1-5 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1

[However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.]

General Electric BWR/6 STS B 2.1.1-6 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) 2.1.1.2b MCPR [ANF Fuel]

The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,

MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the XN-3 critical power correlation. Reference 3 describes the methodology used in determining the MCPR SL.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the XN-3 correlation, the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that there would be no transition boiling in the core during sustained operation at the MCPR SL.

If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.

2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated General Electric BWR/6 STS B 2.1.1-7 Rev. 4.0

TSTF-564, Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria,"

limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. NEDE-24011-P-A, (latest approved revision).

3. XN-NF524(A), Revision 1, November 1983.

4. 10 CFR 100.

General Electric BWR/6 STS B 2.1.1-8 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTIONS LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs), and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated.

Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE ----------------------------------- REVIEWER'S NOTE -------------------------------

SAFETY Incorporate the MCPR95/95 discussion if applicable.

ANALYSES ------------------------------------------------------------------------------------------------

The analytical methods and assumptions used in evaluating the AOOs to establish the operating limit MCPR are presented in the FSAR, Chapters 4, 6, and 15, and References 2, 3, 4, and 5. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (CPR). When the largest CPR is combined with added to the [SL]MCPR[99.9%] SL, the required operating limit MCPR is obtained.

[MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical General Electric BWR/6 STS B 3.2.2-1 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9%

statistical analysis.]

The MCPR operating limits are derived from [the MCPR99.9% value and]

the transient analysis, and are dependent on the operating core flow and power state (MCPRf and MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 3, 4, and 5). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref. 6) and the multichannel thermal hydraulic code (Ref. 7). MCPRf General Electric BWR/6 STS B 3.2.2-2 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 BASES APPLICABLE SAFETY ANALYSES (continued) curves are provided based on the maximum credible flow runout transient for Loop Manual and Non Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control. Non Loop Manual operational modes allow simultaneous runout of both loops because a single controller regulates core flow.

Power dependent MCPR limits (MCPRp) are determined by approved transient analysis modelsthe three dimensional BWR simulator code and the one dimensional transient code (Ref. 8). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The MCPR operating limits specified in the COLR [(MCPR99.9% value, MCPRf values, and MCPRp values)] are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the larger of the MCPRf and MCPRp limits[, which are based on the MCPR99.9% limit specified in the COLR].

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a slow recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses documented in Reference 9 indicate that the nominal value of the initial MCPR expected at 25% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies (Ref. 5) encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor (IRM) provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern.

Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

General Electric BWR/6 STS B 3.2.2-3 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 BASES ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.

Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

B.1 If the MCPR cannot be restored to within the required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is 25% RTP and periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER reaches 25% RTP is acceptable given the large inherent margin to operating limits at low power levels.

[ The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

]

General Electric BWR/6 STS B 3.2.2-4 Rev. 4.0

TSTF-564, Rev. 1 MCPR B 3.2.2 BASES REFERENCES 1. NUREG-0562, June 1979.

2. [Plant specific current cycle safety analysis].

3. FSAR, [Appendix 15B].

4. FSAR, [Appendix 15C].

5. FSAR, [Appendix 15D].

6. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors, Neutronics Methods for Design and Analysis," Volume 1 (as supplemented).

7. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX Thermal Limits Methodology Summary Description," Volume 3, Revision 2, January 1987.

8. XN-NF-79-71(P), "Exxon Nuclear Plant Methodology for Boiling Water Reactors," Revision 2, November 1981.

9. "BWR/6 Generic Rod Withdrawal Error Analysis," General Electric Standard Safety Analysis Report, GESSAR-II, Appendix 15B.

General Electric BWR/6 STS B 3.2.2-5 Rev. 4.0

v t e Letter Sequence Other
CAC:MG0161, (Approved, Closed) EPID:L-2017-PMP-0007, TSTF-564 (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions Answers Results Approval Other: ML18299A069, ML18299A076, TSTF-18-05, TSTF Response to NRC Questions on TSTF-564, Safety Limit (Minimum Critical Power Ratio) MCPR
MONTHYEAR2018-11-16 [Table View]

TSTF-18-05, TSTF Response to NRC Questions on TSTF-564, Safety Limit (Minimum Critical Power Ratio) MCPR

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3.0 REGULATORY ANALYSIS

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