LR-N16-0192, Response to Request for Additional Information Stuck Open Safety/Relief Valves

Response to Request for Additional Information Stuck Open Safety/Relief Valves

ML16348A017
Person / Time
Site:	Hope Creek
Issue date:	12/13/2016
From:	Carr E Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF7709, LR-N16-0192
Download: ML16348A017 (13)

Text

PSEG Nuclear LLC P.O. Box 236, Hancoclts Bridge, New Jersey 08038-0236 PSEG DEC fs 2016 NudearLLC 10 CFR 50.90 LR-N16-0192 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

Response to Request for Additional Information, Re: Stuck Open Safety/Relief Valves (CAC No. MF7709)

Reference:

NRC letter to PSEG, "Hope Creek Nuclear Generating Station- Request for Additional Information Regarding Request to Delete Technical Specification Action Statement 3.4.2.1.b Associated with Stuck Open Safety/Relief Valves (CAC No. MF7709)," dated November 2, 2016 (ADAMS Accession No. ML16271A205)

In the referenced letter, the Nuclear Regulatory Commission (NRC) requested PSEG Nuclear LLC (PSEG) to provide additional information in order to evaluate the proposed deletion of Technical Specification (TS) 3.4.2.1, Action b associated with the subject safety-relief valves (SRVs). Attachment 1 provides a detailed response to the request for additional information. provides revised TS markups based on the responses in Attachment 1. A revision to the 10 CFR 50.92 no significant hazards determination previously submitted is provided in with changes marked with revision bars.

There are no regulatory commitments contained in this letter.

Should you have any questions regarding this submittal, please contact Ms. Tanya Timberman at 856-339-1426.

Page 2 LR-N16-0192 I declare under penalty of perjury that the foregoing is true and correct.

Executed on (Date)

Eric S. Carr Site Vice President Hope Creek Generating Station Attachments:

1. Response to Request for Additional Information

2. Mark-up of Proposed Technical Specification Pages

3. Revised of 10 CFR 50.92 No Significant Hazards Consideration cc: Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE Hope Creek Commitment Tracking Coordinator Corporate Commitment Tracking Coordinator

LR-N16-0192 Attachment 1 Response to Request for Additional Information

LR-N16-0192 Response to Request for Additional Information Request to Delete Technical Specification Action Statement 3.4.2.1.b Associated with Stuck Open Safety/Relief Valves Hope Creek Generating Station Docket Nos. 50-354 By letter dated May 11 2016 (Agencywide Documents Access and Management System Accession No. ML16132A374), PSEG Nuclear LLC (PSEG, the licensee), submitted a license amendment request (LAR) for Hope Creek Generating Station (Hope Creek). The LAR proposes to revise Hope Creek's Technical Specification (TS) requirements by deleting TS Action Statement 3.4.2.1.b concerning stuck open safety/relief valves. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the application and, based upon this review, determined that additional information requested below is needed to complete our review.

TS Action Statement 3.4.2.1.b currently states:

With one or more safety/relief valves stuck open, provided that suppression pool average water temperature is less than 110°F, close the stuck open safety relief valve(s); if unable to close the stuck open valve(s) within 2 minutes or if suppression pool average water temperature is 110°F or greater, place the reactor mode switch in the Shutdown position.

TS 3.6.2.1 Action b.2, also requires the reactor mode switch to be placed in the shutdown position when the suppression chamber average water temperate exceeds 110 degrees Fahrenheit (°F), while in Modes 1, 2, and 3. The licensee states that this action will remain unchanged and provides appropriate remedial action, permitted by TSs, until the limiting condition for operation can be met.

Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, "Technical specifications,"

establishes the NRCs regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

Request for Additional Information In order for the staff to evaluate the licensee's proposed deletion of TS 3.4.2.1, Action b associated with the subject safety-relief valves (SRVs), the NRC staff requests the following additional information:

RAI-1

Since TS requirements are derived from a licensee's licensing basis, please (1) provide documentation/references related to Hope Creek's current licensing basis that requires inclusion of the '2 minute' SRV closure requirement in TS 3.4.2.1.b, and (2) indicate how long this requirement has been in the current TSs.

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LR-N16-0192 PSEG Response to RAI-1

1. In developing the license amendment request, PSEG identified no documentation or references related to Hope Creek's current licensing basis that requires inclusion of the 2 minute SRV closure requirement in TS 3.4.2.1.b. This included review of the Hope Creek TS Bases, Updated Final Safety Analysis Report (UFSAR), and NUREG-1048, "Safety Evaluation Report related to the operation of Hope Creek Generating Station." It was determined that there is no basis for the requirement in these documents. Specifically, the description of operator actions for the inadvertent main steam relief valve opening event in UFSAR Section 15.1.4.2.1.1 does not require the open SRV to be closed within a specific time, only that a shutdown must be initiated if the valve cannot be closed.

2. The current TS action statement 3.4.2.1.b requirement to place the reactor mode switch in the shutdown position if stuck open safety/relief valve(s) are not closed within two minutes has always been in the Hope Creek TS. The original Hope Creek TS were based on NUREG-0123, "Standard Technical Specifications for General Electric Boiling Water Reactors," which also contained the requirement. The 2-minute SRV closure requirement is not discussed in the Bases for either the HC TS or for NUREG-0123.

RAI-2

Please provide additional justification (such as operating experience, etc.) for the deletion of the proposed change.

PSEG Response to RAI-2 In cases where a stuck open SRV is not immediately challenging the TS limit for average suppression pool temperature, the 2 minute TS requirement unnecessarily restricts the operators' opportunity to take actions to reclose the valve and/or to prepare for a plant shutdown.

A search of Licensee Event Reports for plant shutdowns due to stuck open SRV(s) since September 1992, when NUREG-1433, Rev. 0 was issued, identified three instances in which SRVs were stuck open in Mode 1.

Event date Plant LER Event summary 11/09/2008 Brunswick 324/08-002-00 SRV spuriously opened with no operator action Unit 2 or testing in progress. Manual reactor scram was inserted approximately nine minutes after SRV opened based on the suppression pool temperature reaching 109.8°F. SRV closed approximately two seconds after plant shutdown.

09/11/1995 Limerick 352/95-008-00 Manual shutdown in accordance with TS Unit 1 requirements after stuck open SRV could not be closed within two minutes. Suppression pool temperature reached 124°F during the event.

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LR-N16-0192 Event date Plant LER Event summary 01/26/1993 LaSalle 373/93-002-00 Manual scram in accordance with TS Unit 1 requirements after stuck open SRV could not be closed within two minutes during SRV testing.

SRV remained open for approximately 18 minutes. Suppression pool temperature reached approximately 70°F during the event.

TS limit (110°F) was not challenged.

In two instances (Limerick Unit 1 and LaSalle Unit 1), a manual shutdown was initiated after the open SRV could not be closed within two minutes of the control room being made aware that the SRV was open. In one instance (Brunswick Unit 2), the TS requirements allowed additional time to attempt to close the SRV and to reduce reactor power before initiating a manual shutdown. In each instance, the safety consequences of the event were negligible. However, in the case of the Brunswick Unit 2 event, operators had additional time to respond and attempt to reclose the open SRV before being required to initiate a manual scram.

In order for the staff to evaluate the licensee's statement that the requirement for the reactor mode switch to be placed in the shutdown position when the suppression chamber average water temperate exceeds 110 °F while in Modes 1, 2, and 3 will not be changed, the NRC staff requests the following additional information

RAI-3

Please clarify when TS 3.6.2.1 Action b.2 applies. Specifically, address the applicable thermal power.

PSEG Response to RAI-3 As currently presented in the HC TS, TS 3.6.2.1, Action b.2 is read to apply when the suppression chamber average water temperature is greater than 95°F and THERMAL POWER is greater than 1% of RATED THERMAL POWER.

PSEG recognizes the current HC TS allow for ambiguity in implementing TS Action 3.6.2.1 Action b. Therefore, PSEG is proposing to revise TS 3.6.2.1.b to align with the BWR Standard Technical Specifications (NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4," Revision 4, dated April 2012). The proposed changes will remove ambiguity with the TS.

RAI-4

For TS 3.6.2.1 Action b, for a stuck open SRV in Modes 1, 2, or 3, how is the requirement to place the mode switch in shutdown accomplished during the allowed time to restore temperature below the maximum average temperature of 95 °F?

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LR-N16-0192 PSEG Response to RAI-4 In Modes 1 and 2, with THERMAL POWER greater than 1% of RATED THERMAL POWER, TS 3.6.2.1 Action b.2 requires the reactor mode switch to be placed in the Shutdown position, regardless of the time elapsed since suppression chamber water temperature exceeded 95°F. The Hope Creek containment control emergency operating procedure requires operators to runback reactor recirculation flow and to initiate a manual scram before suppression pool temperature exceeds 110°F, regardless of power level.

PSEG recognizes the current HC TS allow for ambiguity in implementing TS Action 3.6.2.1 Action b. Therefore, PSEG is proposing to revise TS 3.6.2.1.b to align with the BWR Standard Technical Specifications (NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4," Revision 4, dated April 2012). The proposed changes will remove ambiguity with the TS.

Attachment 2 provides the existing TS pages marked up to show the proposed changes.

PSEG has determined that the information provided in this submittal does not alter the conclusions reached in the 10 CFR 50.92 no significant hazards consideration (NSHC) determination previously submitted; however, the NSHC is revised to reflect the change to TS 3.6.2.1 Action b (see Attachment 3). In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

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LR-N16-0192 Attachment 2 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License NPF-57 are affected by this change request:

Technical Specification Page 3.6.2.1, Depressurization Systems, Suppression Chamber 3/4 6-12 3.6.2.1, Depressurization Systems, Suppression Chamber 3/4 6-13

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATION

=*=~~==========~==**============a~=a=====~~=~a~=====================~~

3.6.2.1 The suppression chamber shall be OPERABLE with:

a. The pool water:

1. With an indicated water level between 74.5" and 78.5" and a

2. Maximum average temperature of 95°F during OPERATIONAL CONDITION 1 or 2, except that the maximum average temperature may be permitted to increase to:

a) 105°F during testing which adds heat to the suppression chamber.

b) 110°F with THERMAL POWER less than or equal to 1\ of RATED THERMAL POWER.

3. Maximum average temperature of 95°F during OPERATIONAL CONDITION 3, except that the maximum average temperature may be permitted to increase to l20°F with the main steam line isolation valves closed following a scram,

b. A total leakage between the suppression chamber and drywell of less than the equivalent leakage through a l-inch diameter orifice at a differential pressure of 0.80 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. and testing that adds heat to the ACTION: suppression pool is not being performed

a. With the suppression chamber water level outside the above limits, restore the water level to within the limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With the suppression chamber average water temperature greater than 95°F and THERMAL POWER greater than 1\ of RATED THERMAL POWER,

1. restore the average temperature to less than or equal to 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except, as permitted above: .

2. 1. With the suppression chamber average water temperature greater than l05°F during testing which adds heat to the suppression chamber, stop all testing which adds heat to the suppression chamber and restore the average temperature to less than 95° within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. 2. With the suppression chamber average water temperature greater than 110°F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode.

and THERMAL POWER greater than 1% of RATED THERMAL POWER HOPE CREEK 3/4 6-12 Amendment No. 110

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (continued)

ACTION: (Continued)

4. 3. With the suppression chamber average water temperature greater than 120°F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200°F.

SURVEILLANCE REQUIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:

a. By verifying the suppression chamber water volume to be within the limits in accordance with the Surveillance Frequency Control Program.

b. In accordance with the Surveillance Frequency Control Program in OPERATIONAL CONDITION 1 or 2 by verifying the suppression chamber average water temperature to be less than or equal to 95°F, except:

1. At least once per 5 minutes during testing which adds heat to the suppression chamber, by verifying the suppression chamber average water temperature less than or equal to 105°F.

2. At least once per hour when suppression chamber average water temperature is greater than 95°F, by verifying:

a) Suppression chamber average water temperature to be less than or equal to 110°F.

c. At least once per 30 minutes in OPERATIONAL CONDITION 3 following a scram with suppression chamber average water temperature greater than 95°F, by verifying suppression chamber average water temperature less than or equal to 120°F.

d. By an external visual examination of the suppression chamber after safety/relief valve operation with the suppression chamber average water temperature greater than or equal to 17rF and reactor coolant system pressure greater than 100 psig.

e. In accordance with the Surveillance Frequency Control Program by a visual inspection of the accessible interior and exterior of the suppression chamber.

HOPE CREEK 3/4 6-13 Amendment No. 193

LR-N16-0192 Attachment 3 Revised of 10 CFR 50.92 No Significant Hazards Consideration

LR-N16-0192

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration PSEG 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed TS change deletes Action Statement 3.4.2.1.b concerning safety/relief valves and revises TS Action Statement 3.6.2.1.b to be consistent with the BWR Standard Technical Specifications (NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4," Revision 4, dated April 2012). The two (2) minute action represents detailed methods of responding to an event, and therefore, if eliminated, would not result in increasing the probability of the event, nor act as an initiator of an event. Limiting condition for operation 3.6.2.1, "Depressurization Systems - Suppression Chamber," and plant procedures provide operators with appropriate direction for response to a suppression pool high temperature (which could be caused by a stuck open relief valve). Providing specific direction to close the valve within two (2) minutes does not provide additional plant protection beyond what is provided for in plant procedures and TS 3.6.2.1.

Therefore, this action can be eliminated, and will not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed TS change deletes Action Statement 3.4.2.1.b concerning safety/relief valves and revises TS Action Statement 3.6.2.1.b to be consistent with the BWR Standard Technical Specifications (NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4," Revision 4, dated April 2012). This change does not change the design or configuration of the plant. No new operation or failure modes are created, nor is a system-level failure mode created that is different than those that already exist.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No 1

LR-N16-0192 The proposed change does not involve a significant reduction in a margin of safety, nor does it affect any analytical limits. There are no changes to accident or transient core thermal hydraulic conditions, or fuel or reactor coolant boundary design limits, as a result of the proposed change. The proposed change will not alter the assumptions or results of the analysis contained in the Updated Final Safety Analysis Report (UFSAR).

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

Based upon the above, PSEG 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.

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