ML20197H337

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Proposed Tech Specs Revising SLMCPR from 1.07 to 1.09 for Two Recirculation Loop Operation & from 1.08 to 1.10 for Single Loop Operation
ML20197H337
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 12/15/1997
From:
NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML19317C821 List:
References
NUDOCS 9712310251
Download: ML20197H337 (25)
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Text

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ATTACHMENT A NIAGARA MOHAWK POWER CORPORATION '

LICENSE NO. NPF-69.

DOCKET NO. 50 410 2

Proposed Channe to Technical Specifications Replace existing pa0&s lil,21 and 3/4 4-1 with the attached revised pages. Replace existing Bases pages 821,82 2, B2 3 and B2 4 with the attached revised pages. These pages have been ratyped in their entirety with marginal markings to indicate changes to the text.

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9712310251 971215 PDR ADOCK 05000410 P PDR ,

.: INDEX DEFINITIONS

  • PAGE

..1.48 VENTILATION EXHAUST TREATMENT SYSTEM ..,.... ................. 1-9.

1.49 V E N TI N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 9

.1.50. COR E OPERATING ' LIMITS REPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 9 -

Table 1.1' Surveillance Frequency Notations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-10 Table 1.2 Operational Conditions . . . . . . . . . . . . . . . . . . . . .; . . . . . .. . . . . . . . . . . . . 1 -11 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS

. THERMAL POWER, Low Pressure or Low Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 .

THERMAL POWER, High Pressure and High Flow . . . . . . . . . . . . . . . . . . . . . . . . . 21 ~ .

Reactor Coolant System Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 Reactor Vessel Water Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints . . . . . . . . . . . . . . . . . . . . . . 2 2 Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints . . . . . . . . . . . . . . . . 2-3 BASES FOR SECTION 2.0 L

2.1 SAFETY LIMITS Introduction ...................................................B2-1

' THERMAL POWER, Low Pressure or Low Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . B2-1 THERMAL POWER, High Pressure and High Flow- . . . . . . . . . . . . . . . . . . . . . . . . . B2-2 Bases Table B2.1.2-1 D ele t e d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-3 Bases Table 82.1.2-2 D e l e t e d . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2 4 NINE MILE POINT - UNIT 2 iii - Amen'dment No. //

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.- 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or Low Flow' ,

2.1.1 -THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the

. reactor vessel steam dome' pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor .

vessel' steam dome pressure less than 785 psig or core flow less than 10% of rated flow,  ;

be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.

THERMAL POWER. Hiah Pressure and Hioh Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09 with two recirculation loop operation and shall not be less than 1.10 with single recirculation

' loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With MCPR less than 1.09, with two recirculation loop operation or less than 1.10 with l single loop operation,- the reactor vessel steam dome pressure greater than 785 psig, and

. core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessei steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1,2,3, and 4.

ACTION:

With the reactor coolant system pressure as measured in the reactor vessel steam dome above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

NINE MILE POINT - UNIT 2: 2-1 Amendment No. /d

e .. . _. _ _

3/4.4- REACTOR COOLANT SYSTEM

"' 3/4.4.1 RECIRCULATION SYSTEM

- RECIRCULATION LO0fS LIMITING CONDITIONS FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:

a. Total core flow greater than or equal to 45% of rated core flow, or
b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1.

APPLICABILITY: OPERATIONAL OONDITIONS 1* AND 2*,

ACTION:

a. With one reactor coolant system recirculation loop not in operation:
1. Within four hours:

a) Place the recirculation flow control system in the Loop Manual (Position Control) mode, and b) Reduce THERMAL POWER to $70% of RATED THERMAL POWER, and, c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.10 per Specification 2.1.2, and, l d) Reduce the Maximum Average Planar Linear Heat Generation Rote (MAPLHGR) limit per Specification 3.2.1, and, e) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specifications 2.2.1, 3.2.2 and 3.3.6.

f) Reduce the volumetric drive flow rate of the operating recirculation loop to s 41,800 *

  • gpm.
  • . See Special Test Exception 3.10.4.
    • T'.iis value represents the volumetric recirculation loop drive flow which produces 100%

core flow at 100%-THERMAL POWER.

I NINE MILE POINT - UNIT 2 ' 3/4 4-1 Amendment No.'/d, //

Cor : ted July 17,1990

. 2.1 BASES FOR SAFETY LIMITS 2.

1.0 INTRODUCTION

The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the_ integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set so that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit so that the MCPR is

'not less than 1.09 for two recirculation loop operation and 1.10 for single recirculation loop operation. MCPR greater than 1.09 for two recirculation loop operation and 1.10 for single recirculation loop operation 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 so'.rce is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses that occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. Although 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 Safety Limit is defined with a margin to the conditions that would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER. Low Pressure or Low Flow The use of critical power correlations is not valid for all critical power calculations performed at reduced pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. 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 greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10 lb/hr, 3 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 l head will be greater than 28 x 10 310/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 of more than 50% of RATED THERMAL PO'WER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

NINE MILE POINT - UNIT 2 B21 Amendment No. /d, /d

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. BASES FOR SAFETY LIMITS 2.1.2 THERMAL PCWER. Hioh Pressure and Hiah Flow The fuel cladding integrity Safety Limit is set so that no 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 resulting in a departure from nucleate boiling have been used to mark the beginning of the region in

. which fuel damage could occur, Although it is recognized that a departure from nucleate 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.

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 Safety Limit is defined as the CPR 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 cistribution within the core and all '

uncertainties.

The Safety Limit MCPR is determined using a statistical model that combines all of the

. uncertalntles in operating parameters and the procedures used to calculate critical power.

The probability of the occurrence of boiling transition is determined using an approved critical power correlation. The critical power correlction is valid over the range of conditions used in the tests of the data used to develop the correlation. Details of the fuel cladding integrity Safety Limit calculation are given in Reference 1. Reference 1 also includes a tabulation of the uncertair. ties used in the determination of the Safety Limit MCPR. The plant specific values of the parameters used in the Safety Limit MCPR statistical analysis are found in the cycle specific analysis.

References:

1. General Electric Standard Application for Reactor Fuel, NEDE 24011-P-A (latest approved revision).

NINE MILE POINT - UNIT 2 B2-2 Amendment No. dd

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.: NINE MILE POINT UNIT 2 82-4' Amendment No. dd, dd

o ATTACHMENTB 3 NIAGARA MOHAWK POWER CORPORATION LICENSE NO. NPF-69 '

DOCKET NO. 50 410 Suncortina information and No Slanificant Hazards Consideration Analysis INTRODUCTION On May 24,1996, GE notified the NRC of a reportable condition involving the generic -

safety limit calculational methodology. GE determined that the generic Safety Limit MCPR -

might be non-conservative when applied to some cycle specific core and fuel designs. As a result of this error, GE performed a cycle specific Safety Limit calculation. NMPC submitted NMP2 LER 96-06 on June 3,1996, and provided additional information regarding the impact of the nonconservative values. NMPC concluded that neither the Safety Limit nor the Operating Limit would have been exceeded for any analyzed plant transient, based on the increased Safety Limit value and the core performance up to that point in the operating cycle. The Supplemental Reload Licensing Report, USAR and COLR have been revised to include the correct Safety Limit Minimum Critical Power Ratio (MCPR) for the current operating cycle. This change to the TS completes the corrective actions described in LER 96-06. NMPC and GE have completed an analysis to determine the necessary Safety Limit MCPR for the upcoming operating cycle (Cycle 7). This analysis shows that the required Safety Limit MCPR will decrease from the current cycle's value based on the cycle specific Safety Limit MCPR calculation. As a result, this TS change revises the Safety Limit MCPR from 1.07 to 1.09 for two recirculation loop operation and from 1.08 to 1.10 for single loop operation to account for these changes.

A cycle specific Safety Limit calculation will be performed for future core reloads using cycle specific core loading patterns and power distributions, pending a long term solution.

Thus, each core reload will be evaluated to ensure that the Safety Limit MCPR conservatively bounds the respective reload and operating cycle. NMPC will utilize its administrative control process to ensure that a cycle specific analysis has been completed prior to each startup from a iefueling outage. This practice will be maintained as long as cycle specific calculations are required.

- In addition, Bases Section 2.1 is revised for consistency on page B2-1 and to delete some of the detailed information and refer instead to General Electtc i Standard Application for Reactor Fuel (GESTAR 11), NEDE-24011-P-A and to the cycle specific analysis (Supplemental Reload Licensing Report). This level of detail is consistent with that in the improved Standard Technical Specifications (NUREG-1434). Some of the wording has been revised to more accurately reflect current practices.

A fnotnote on page 3/4 41 is also deleted. This footnote only applied to the first operating cycle and is therefore no longer applicable or neceasary.

Page 1 cf 4

o^

l ANALYSIS The reload analyses and evaluations are performed based on General Electric Standard Application for Reactor Fuel, NEDE 24011 P-A 13 and NEDE 24011-P-A-13-US (GESTAR 11, latest approved revisions). This document describes the fuellicensing acceptance criteria; the fuel thermal-mechanical, nuclear, and thermal hydraulic analyses bases: and the safety analysis methodology. The evaluations included transients and accidents likely ,

to limit operation because of MCPR considerations, overpressurization events, loss of coolant accident, and stability analysis.

Core operating limits are established to ensure that the Safety Limits are not exceeded, the limits of 10CFR50.46 are satisfied, and that other fuellicensing acceptance criteria specified in GESTAR 11 are met. The fuel cladding is one of the principal barriers to the

. release of radioactive materials to the environment. Safety Limits are established to protect the integrity of this barrier during normal plant operation and anticipated transients.

The Safety Limit MCPR is applied to ensure fuel cladding integrity is not lost as a result of over-heating. Compliance with the Safety Limit MCPR will assure that 99.9 percent of the fuel rods would not be expected to experience transition boiling during the most limiting anticipated operational occurrence.

Consistent with the bas:s of GESTAR 11, the limiting transient events are reanalyzed for each reload. These analyses include those transients which could result in a significant reduction in MCPR. As a result of GE's work on cycle specific safety limits, a nonconservative Safety Limit MCPR was identified and a Part 21 notification was made to the NRC. As documented in the Supplemental Reload Licensing Report for Nine Mile Point Nuclear Power Station Unit 2, GENE 24A5174, Rev.1, June 1996, the Safety Limit MCPR was increased from 1.07 for two recirculation loep operation and 1.08 for single loop operation to 1.10 and 1.12 respectively. This change was implemented in the COLR via an increase in the Operating Limit MCPR to maintain the existing margin of safety. The COLR contains the cycle specific parameters that were removed from the TS in TS Amendment No.17, dated June 19,1990.

The upcoming operating cycle (Cycle 7) has been analyzed in accordance with the NRC approved methods (described in GESTAR 11) and subsequent commitments made in GE's letter to the NRC dated May-24,- 1996, regarding the 10CFR Part 21 reportable condition relating to the Safety Limit MCPR calculation. The methodology used for calculating the

. cycle specific Safety Limit MCPR is as described in Amendment 25 to GESTAR 11, which was submitted by GE to the NRC on December 13,1996. Additional information can be found in Attachment D. As a result of the cycle specific calculation, the Safety Limit MCPR for Cycle 7 will be 1.09 for two recirculation loop operation and 1.10 for single loop operation. Thus, this TS change revias the Safety Limit MCPR to 1.09 for two recirculation loop operation and 1.1' ,or single loop operation to account for these changes.

The deletion of the TS Bases tables will not affect the methodology used to calculate the fuel cladding safety limit. As described in the proposed wording in the Bases, GESTAR 11 includes the appropriate uncertainties. The nominal va!ues are replaced with cycle specific Page 2 of 4

l .

I values, which are contained in the cycic specific analysis (Supplemental Reload Licensing Report). Thus, the NRC approved methods will still be used to calculate the safety limit using the appropriate input.

The deletion of the footnote is an administrative change only, and has no impact on plant operation. This footnote only applied to the first operating cycle and is therefore no longer necessary.

CONCLUSIOliS A cycle specific safety limit calculation was performed for NMP2. The methodologies used to determine the limits were based on those approvec by the NRC in GESTAR 11 and subsequent commitments made in GE's letter to the NRC dated May 24,1996, regarding the 10CFR Part 21 reportable condition. These methodologies will continue to assure that the fuellicensing acceptance criteria are met. Based on the evaluations and analyses o'escribed, NMP2 can be safely operated with the revised Safety Limit MCPR.

HQ_SIGnllFICANT HAZARDS CONSlpERATION ANALYSIS 10CFR50.91 requires that at the time a licensee requests an amendment, it must provide to the Commission its analysis, using the standards in 10CFR50.92 concerning the issue of no significant hazards consideration. Therefore, in accordance with 10CFR50.91, the following analysis has been performed with respect to the requested change.

The operation of Nine Mile Point Unit 2. in accordance with the orooosed amendment, will not involve a sionificant increase in the orobability or consecuences of an accident DLeviousiv evaluated.

The derivation of the revis3d Safety Limit MCPR was performed using the NRC approved methodology in GESTAR 11. The Sefety Limit MCPR is a TS numerical value that cannot initiate an event. Maintaining compliance with this limit will assure that 99.9 percent of the fuel rods will not experience transition boiling during transient events. The deletion of the footnote that is no longer necessary and the revision to the Bases information are administrative only. The proposed change does not modify any of the accident initiators described in the USAR. No equipment malfunctions or procedural errors are created as a result of this enange, therefore, no accidents are afiected by it. The change does not adversoly impact the integrity of the fuel cladding, which is the first barrier to the release of radioactivity to the environment. The change does not affcct the operation of any systems necessary to mitigate the radiological consequences of an accident or to safely shutdown the plant. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previcusly evaluatad.

Ibteneration of Nine Mile Point Uait 2. in 3ccordhDEe with.the orocosed amendment, wlH not create the oossibl4tv of a new or different kind of accident from any accident DI9viousiv evaluated.

The Safety Limit MCPR is a TS numerical value designed to prevent fuel damage from transition boiling. It cannot create the possibility of a transient or accident. The deletion of the footnote that is no longer necessary and the revision to the Bases information are administrative only. The proposed change does not directly impact the operation of any Page 3 of 4

l systems or equipment important _to safety. The analyses show that all fuel licensing _

acceptance criteria are met. The fuel cladding, reactor vessel, and reactor coolant system integrity will be maintained. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

Ibe onoration of Nine Mile Point Unit L in accordance with the oronosed amendment will attinvolve a sianificant reduction in a marain of safety. ,

The Safety Limit MCPR calculation was performed using the NRC approved methodology in GESTAR ll. Analyses of limiting USAR transients establish Operating Limit MCPR values that ensure that the Safety Limit MCPR is not violated. The revised cycle specific Safety Limit MCPR preserves the existing margin of safety and will continue to assure that 99.9 percent of the fuel rods will not experience transition boiling during transient events. The deletion of the footnote that is no longer necessary and the revision to the Bases information are administrative only. Thus, the margin of safety to fuel cladding failure due to insufficient cladding heat transfer during transient events is not reduced. Therefore, this change will not involve a significant reduction in a margin of safety.

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ATTACHMENT C NIAGARA MOHAWK POWER CORPORATION LICENSE NO. NPF 69 DOCKET NO. 50-410 Marked-Un Cony of Proposed Channe to Current Technical Snecifications The current version of NMP2 Technical Specificati- pages lil, 2-1, 3/4 4-1, B21, B2 2, 82 3 and B2-4 have been hand marked-up to reflect the proposed changes.

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!NDEX .

( DEFINITIONS PAGE 1.48 VENTILATION EXHAUST. TREATMENT SYSTEM.........................

1-9 1.49 VENTING............................................. ........ 1-9 1.50 CORE OPERATING LIMITS REP 0RT................................. 1-9 Surveillance frequency 1-10 Table 1.1 Notations........................

Table 1.2 Operational Conditions......., ......................... 1-11 2.0 SAFETY-tIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or low F1ow....................... 2-1 THERMAL POWER, High Pressure and High F1ow.................... 2-1 Reactor Coolant System Pressure............................... 2-1 2-1 Reactor Vessel Water Level....................................

. 2.2 LIMEINGSAFETYSYSTEMSETT!sGS Reactor Protection System Instrumentation Setpoints........... 2-2 Table 2.2.1-1 Reactor Protectton System Instrumentation Setpoints. 2-3 BASES FOR SECTION 2.0 i

l 2.1 SAFETY LIMITS -

l Introduction..................... ............................ 82-1 l

82-1 THERMAL POWER. Low Pressure or Low flow.......................

THERMAL POWER, High Pressure and High Flow.................... 82-2

.. 82-3

'l es o rame n Bases Tabl B2-4

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< . NINE MILE POINT - UNIT 2' 111 AmendmentNo.//

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. 2' . 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS I

2.1 SAFETYLi'~ITSM )

THERMAL POWER, Low Pressure or low Flow l 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the l reactor vessel steam dome pressure less than 785 psig or core flow less than '

10% of rated flow.

i APPLICABILIT_Y: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam done pressure less than 785 psig or core flow less than 10% of rated

' flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the require-ments of Specification 6.7. .

. THERMAL POWER. High Pressure and High Flow

\l.09 2.1.2 - The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be thanJhef with two recirculation loop operation and shall not be less than with s ngle recirculation loop operation with the reactor vessel steam pressure greater than 785 psig and core tiow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CON 0!TIONS 1 and 2.

i ACTION: l I 09 MCPR less than . , with two recirculation loop operation or less than  !

t.f o with single loop operation, the reactor vessel steam done pressure greater an 785 psig, and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7. ,

l REACTOR C00LANT SYSTEM PRESSURE 2.1.3 The reactor coolant system prassure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

l With the reactor coolant system pressure as measured in the reactor vessel  !

steam dome above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant j system pressure less than or equal to 1325 p;;ig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and c,ayly with the requirements of Specification 6.7.

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel, t

NINE MILE POINT - UNIT 2 2-1 AmendmentNo.J7 l

,=. .

3/4.4- REACTOR'C00LANT_ SYSTEM 3/4.4.1

  • RECIRCULATION SYSTEM

-RECIRCULATION LOOPS

-LIMITING CON 0!TIONS FOR OPERATION *

- 3.4.1.1 - Two reactor coolant system recirculation loops shall be in operation with:

a.- . Total core flow greater than or equal to 45% of rated core flow, or

b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With one reactor coolant system recirculation loop not In operation: '

1.- Within four hours:

a) Place the recirculatica flow control system in the Loop

  • Manual (Position Control) mode, and-b) Reduce. THERMAL POWER to 1 70% of RATED THERMAL POWER, and, c) Increase the MIN ICAL F0WER RATI? (MCPR) Safety Lim'it by 0.01 to per Specification 2.1.2, and, 0

d) Reduce the Maximum verage Planar Linear Heat Generation Rate f t

(MAPLHGR) limit per Specification 3.2.1, and, e) Reduce- the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those appitcable for single recirculation loop operation per-Specifications 2.2.1, 3.2.2 and 3.3.6.

f) Reduce the volumetric drive flow rate of the operating recirculation loop to 1 41,800** gpm.

  • See Special Test Exception 3.10.4.
    • This' value represents the volumetric recir:ulation loop drive flow which produces 100% ' core flow at 100% THERMAL POWER.

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+yc,b NINE-MILE POINT - UNIT 2 3/4 4-1 AmendmentNo.16,)@#

-Ectrected Jul- f 17, 199F i .

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( - 2.1 BASES FOR SAFETY LIMITS 2l1LO INTRODUCTION The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are ,

established to protect the integrity of these barriers during normal plant operations and

- anticipated transients. The fuel cladding integrity Safety Limit is set so that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a.a - ek spproach is used to establish a Safety L mit so that t or two recirculation operatinn and or singio /,/0 OI.04 ~~MCPM is not less .

the ration. MCPR greater than _ for two recirc a son loor ,ay recirculation lo I peration a or single recirculation loop operatibn represents a conservativ 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 that occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. Although fission .

! product migration from cladding perforation is }ust 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 Safety Limit is defined with a margin to the

' contlitions that would produce onset of transition boiling, MCPR of 1.0. These

( conditions represent a significant' departure from the condition intended by design for planned operation.

21.1 THERMAL POWER. Low Pressure or low Flow f

The use of critical power correlations is not valid for all critical power calculations O i ,

performed at reduced pressures below 785 psig or core flows less v tha This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass reg;on is essentially all elevation head, the core pressure drop at low power and flows will 5 always be greater than 4.5

' psi. Analyses show that with a bundle flow of 28 x 10 lb/hr, bundle pressure' drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow t

with a 4.5Jhsi driving head will be greater than 28 x 108 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 of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

4 i '

4 NINE MILE POINT - UNIT 2 B21 Amendment No. )6 ff(

1

- _ _ . . .___w_____-___-__.__. . . . . - - ..:-.~ _ . _ -- , - - ,

- .- - .- - .~--- . - . - . . - - . ,

~ BASES FOR SAFETY LIMITS 2.1.2 THERMAL POWER. Hioh Pressure and Hiah Flow The fuel cladding integrity Safety Limit is set'so that no fuel damage is calculated to-occur if the limit it, not violated. Since the parameters that result in fuel damage are not directly observable during reactor operationi the thermal and hydraulic conditions  ;

resulting in a departua from nucleate boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to SWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. 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 Safety Limit is defined as the CPR in the limiting fuel assemb'y 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 Safety Limit MCPR is determined using a statistical model that combines all of the j uncertainties in operating parameters and the procedures used to calculate entical power. The probability of the occurrence of boiling transition is determined using an approved critical power correlation. The critical power correlation is valid over the range'U of conditions used in the tests of the data used to develop the correlation. t

( The reedited inpu the stati ' al model the uncert ties listed

  • Bases T le nominal v es of the e paramete listed in B es Table .1.2 2.

i B2. dt 1 and

( bases the uncert ' ties in the e paramet and the b is for the certa. int n the cr' al power c elation are en in Refer ce 1. The wer distririly tion is chos to base a typical 4 assembly te in which rod patte was arbi pr uce a skew power distri tion having greatest n ber of asp 6mblies a he ghost power vols. The rat distribution during any el cycle wpdid not be s severa as t distribution ed in the anafysis.

References:

1. General Electric Standard Application for Reactor Fuel, MEDE 24011 P A (latest -

approved revision).

V ZNSGRT* A e

k NINE MILE POINT - UNIT 2 B2 2 Amendment No. h

. . , ,*=

INSERT A - .;

Details of the fuel cladding integrity Safety Limit calculation are given in Reference 1. _._ l Reference 1 also includes a tabulation of the uncertainties used in the determination of the  ;

Safety Limit MCPR. The plant specific values of the parameters used in the Safety Limit :

MCPR statistical ana!ysis are found in the cycle specific analysis; ,

h h

-.- . - - - . - -- ..n.,

f CASES TABLE B2.1.21 UNCERTAINTIE. ED IN THE DETERMINATION OF THE FMEL CLADDING SAFETY LIMIT

  • STANDARD-DEVIATION

(% OF POINT)

QUANTITY 1.76 Feedwater Flow 0.76 Feedweter Tem reture 0.5 Reactor Prep re 0.2 Core ini emperature 2.5 /l Core otal Flow' 3.0 annel Flow Area 10.

Friction Factor Multiplier

.O Channel Friction Factor Multiplier 8.7 l TIP Readings 1.6 R Factor 3.0 l Critical Power

/

  • The 'ncertainty analysis used to establish t corewide Safety Limit MCPR is d on the asrumption of quadrant pow symmetry for the reactor core. The on loop operation and single l ues herein apply to both two recircut y recirculation loop operation. l J

i B2-3 Amendment No.J4f 8 <

NINE MILE POINT - UNIT 2

h* SASES TABLE B2.1.2 2 NOMINAL V S OF PARAMETERSo USED IN .

THE STATISTICAL ANALYEfS OF FUEL CLADDING INTEGRITY SAFETY LIMIT ** l s

EMt ETER VALUE ERMAL POWER 3293 MW

. Core Flow . 102.5 /hr .

Dome Pressure 100 psig '

Bundle Enrichment .0 Wt % U 235 R F;ctor:

0 - 10 GWD/ST 0.915 10 - 15 GWD/ 0.954

> 15 GW ST 0.954 6,

  • The ues in this table are for a representative plant l
    • e Statistical analysis has been evaluated an own to be valid at 3467 MW(t) with GE i

fuel (

References:

NEDC 31984P, " Generic atuations of GE BWR Power Uprate", Volume 1: NEOC-24011 P A, GESTAR 11; and N -31152P. "GE Fuel Bundle Designs").

l3 C >

A l

NINE MILE POINT - UNIT 2- B2-4 Amendment No. d5 [f

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

) **_' '

l T d.. , ,

ATTACHMENT F NIAGARA MOHAWi; POWER CORPORATION LICENSE NO. NPF 69 DOCKET NO. 50-410 Non-Proprietary Version of Information Renarrana GE's l Methodolony for Cycle Specific Safety Limit MCPR Calculations ATTACHED

l. .A t 2 J .

Attachment Additional Information Regarding the 1,09 December 2,1997 Cycle Specific SLMCPR for Nine Mile Point-2 Cycle 7 References

[ l] General Electric BWR Thermal Analysis Bmis (GETAB): Data, Correlation and Design Application, NEDO-10958 A, January 1977.

[ 2) General Electric Standard Applicationfor Peactor Fuel (GESTAR 11), NEDE-24011 P. A-11, November 1995.

[ 3] General Electric Standard Applicationfor Reactor Fuel (GESTAR lb, NEDE-240 l 1-P-A-l 3, August 1996.

[ 4) General Electric Fuel Bundle Designs, NEDE-31152-P, Revision 6, April 1997.

[ $) Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations, N EDC-32601 P, Class IIl, December 1996.

[ 6) R Factor Calculation Methodfor Gell, gel 2 and GE13 Fuel. NEDC 32505P, November 1995.

Comparison of NMP-2 Cycle 7 SLMCPR versus the Cycle 6 SLMCPR Table i summarizes the relevant input parameters and results of the SLMCPR determination for both the NMP-2 Cycle 6 and Cycle 7 cores. Both cycle-specific evaluations were performed using the O

V methods described in GETABl ll. The evaluations yield different calculeted SLMCPR values because the inputs that are used are .Gfferent. The quantities that have been sl. awn to have some impact on the determination of the safety limit htCPR (SLMCPR) are provided, hiuch of this information is redundant but is provided in this case because it has been provided previously to the NRC to assist them in understanding the differences between plant / cycle specific SLhiCPR evaluations. (())

Prior to 1996, GESTAR 11[ 2] stipulated that the SLMCPR analysis for a new fuel design be performed for a large high power density plant assuming a bounding equilibrium core. The gel 1 product line generic SLMCPR value of 1.07 was determined according to this speciGcation. Later revisions to GESTAR 11[ 31 have been submitted to the NRC to describe how plant / cycle specific SLMCPR analyses are used to conGrm the calculated SLMCPR value on a plant / cycle speciGc basis using the uncertainties denned in Reference [ 4].

In comparing the NMP-2 Cycle 6 and Cycle 7 SLMCPR values it is important to note that although gel 1 dominates both cores, neither core is an equilibrium core. C6 was a mixed core with gel 1 and GE9 fuel. In both cores, the fresh bundle enrichment increases relative to the previous cycle fresh fuel. The freshest fuel is the latest batch of gel I that comprises (( )) of the bundles in the core. Also, this fresh Mch of GElI has the highest enrichment (( )), as compared to a core average enrichment of(( )), a.,;Jaown in Table 1. By way of comparison, the Cycle 6 core has a smaller batch of fresh gel 1 (( )) and a lower core average enrichment of(( )). liigher enrichment in the fresh fuel for the NMP-2 Cycle 7 core (compared to the rest of the core) produces higher power in the fresh bundles relative to the rest of the core. These enrichment differences result in the gel i q fresh fuel producing a higher relative share of the number of fut rods calculated to be susceptible to boiling transition (NRSBT). (( ))

(( GENE Proprietary information )) page 1 of 3 Ib

(( enclosed by double brackets ))

rg .*

i 4

Attachment Additional Information Regarding the 1.09 December 2,1997 O Cycle Specific SI,MCPR for Nhic Mile Point 2 Cycle 7

(( )) The NMP 2 Cycle 7 core has a somerhat Oatter core MCPR distribt.Lon than the Cycle 6 core and the fresh Cycle 7 bundle R factor distributions are somewhat more peaked than the fresh Cycle 6 bundles.

(( ))

[I 11

(( ]l one is led to the conclusion that the tore MCPR distribution for NMP 2 Cycle 7 is slightly natter than the distribution evaluated for Cycle 6.

The uncontrolled bundle pin by pin power distributions were compared between the NMP 2 Cycle 6 and Cycle 7 bundles used for the SLMCPR analyses. Pin-by-pin power distributions are characterlied in terms of R factors using the methodology denneiin Reference [ 6]. For the NMP 2 Cycle 7 bundles, there is a sornewhat more peaked distribution of uncontrolled R factors for the highest power rods in each bundle, which in the calculation are the rods most likely to be susceptible to boiling transition. This fact is suggested by a graphical comparison of the bundle R factor distributions but is dif0 cult to quantify graphically since the relative flatnesses are similar and the rods that have an R factor closer to the R factor for the lead rod are statistically worth much more than those that have R-factors that are further away. (( ))

The Ontness of the pin R factor distribution within a particular bundle is characterized (( ))

(( )) Thus this supports the conclusion that the lower St.MCPR value for NMP 2 Cycle 7 is at least in part due to the more peaked bundle R factors relative to those used for the Cycle 6 St.MCPR evaluation.

O

(( GliNE Proprietary information )) page 2 of 3 ,

l[ enclosed by double brackets ))

g .*

-)

December 2,1997

( Additional Information Regarding the 1.09 Attachment Cycle Specific SLMCPR for Nine Mile Point-2 Cycle 7 "I able 1 Comparison of NMP 2 Cycle 6 and Cycle 7 Core and Hundle Quantitles that impset the SLMCPk ll Summary The calcult.ed nominal 1.09 Monte Carlo SLMCPR for NMP 2 Cycle 7 is consistent with what one would expect (( )) the I 09 SLMCPR value is appropriate. Various quantities (( }l have been used over the last year to compare quantities that impact the  ; calculated SLMCPR value. These other quantities have been provided to the NRC previously for j other plundycle specific analyses using a fonnat such as that given in Table 1. These other , quantities have also been compared for this core / cycle (( )). The key parameters in Table 1 support the conclusions that the NMP 2 Cycle 7 core / cycle has a slightly flatter core MCPR distnbution and somewhat more peaked in bundle power distributions (( )) than what was used to perfonn the Cycle 6 SLMCPR evaluation. The more peaked bundle R factor distribution slightly outweighs the flatter core MCPR distribution, resulting in the lower calculated Cycle 7 SLMCPR relative to the Cycle 6 SLMCPR evaluation. Ilased on all of the facts, observations and arguments presented above, it is concluded that the i calculated SLMCPR value of 1.09 for the NMP 2 Cycle 7 core is appropriate, it is reasonable that this value is 0.01 lower than the 1.10 value calculated for Cycle 6. l'or single loop operations (SLO) the safety limit MCPR is 0.01 greater than the two loop value (( )) 

 ]

(( GENE Prorricta y Information )) page 3 of 3 [{ cnclosed by double brackets )) 

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