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Nuclear Business Unit Hancocks Bridge, New Jersey 08038-0236@
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('. /7?/m:Y) Public Service Electric and Gas Company
                                                                                                                      , /Yo-y ~ /777 Public Service Electric and Gas Company
* P.IT. Box 236 JUL 0 7 1997 LR-N970390 Chief, Rules Review and Directives Branch Division of Administrative Services Off ice of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 SALEM GENERATING STATION UNITS 1 AND 2 HOPE CREEK GENERATING STATION :0 ::0
* P.IT. Box 236 Hancocks Bridge, New Jersey 08038-0236@
::D v.; tt{i cs;:o .-("') (/)0 .LSJ :IJ* Oco SC ""'""' FACILITY OPERATING LICENSES.
Nuclear Business Unit JUL 0 7 1997 LR-N970390 Chief, Rules Review and Directives Branch Division of Administrative Services Off ice of Administration                                                                                       :0  ::0 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
DPR-70 I DPR-75 AND NPF-57 rn -..o* DOCKET NOS. 50-272, 50-311 AND 50-354 Z *** CJ COMMENTS ON NUREG-1606, "PROPOSED REGULATORY GUIDANCE IMPLEMENTATION OF 10 CFR 50. 59 (CHANGES, TESTS, OR EXPERIMENTS)" Gentlemen:
                                                                                                                                ~ ~c:::~ ::D v.;       tt{i cs;:o   .-   ("')
Public Service Electric and Gas Co. (PSE&G) provides the attached comments in response to the Federal Register notice dated May 7, 1997. Our comments are directed towards those sections of the NUREG which we believe to be new requirements.
(/)0 SALEM GENERATING STATION UNITS 1 AND 2                                                                      .LSJ HOPE CREEK GENERATING STATION                                                                                :IJ*         ~
Their effect would be to greatly increase the number of proposed changes requiring NRC review and approval with little or no safety benefit. In addition to the detailed comments attached to this letter, PSE&G also endorses the comments submitted by the Nuclear Energy Institute and by the Licensing and Design Bases Clearinghouse.
FACILITY OPERATING LICENSES. DPR-70 I DPR-75 AND NPF-57                                                     Oco-~*J SC    ""'""'
PSE&G appreciates the opportunity to provide comments on the draft document.
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Should there be any questions concerning this submittal, please do not hesitate to contact us. Sincerely, (,.--------
                                                                                                                                '}~  -..o*
-9707150120 970707 David R. Powell Manager-' PDR NUREG 1606 C PDR Attachment (1) The power is in your hands . ... -f"\{'\.i:*1 ! ' ., 1 _. ... .
DOCKET NOS. 50-272, 50-311 AND 50-354                                                                           Z *** CJ COMMENTS ON NUREG-1606, "PROPOSED REGULATORY GUIDANCE REubf.ED~*o IMPLEMENTATION OF 10 CFR 50. 59 (CHANGES, TESTS, OR EXPERIMENTS)"
-L i cens ing and Regulation
Gentlemen:
-* I 1\1111 \Ill\ \Ill\ \11\IJll\\
Public Service Electric and Gas Co. (PSE&G) provides the attached comments in response to the Federal Register notice dated May 7, 1997.                                                                                                                                         I Our comments are directed towards those sections of the NUREG which we believe to be new requirements.                                         Their effect would be to greatly increase the number of proposed changes requiring NRC review and approval with little or no safety benefit.
\Ill\ \1111\11 : 95-2168 REV. 6/94 I 
In addition to the detailed comments attached to this letter, PSE&G also endorses the comments submitted by the Nuclear Energy Institute and by the Licensing and Design Bases Clearinghouse.
*.
PSE&G appreciates the opportunity to provide comments on the draft document.               Should there be any questions concerning this submittal, please do not hesitate to contact us.
* Document Control Desk LR-N970390
Sincerely,
* C Mr. H. J. Miller, Administrator  
(,.-------- -9707150120 970707                                   David R. Powell
-Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 JUL 0*7 1997 Mr. L. Olshan, Licensing Project Manager -Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. D. Jaffe, Licensing Project Manager -Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. S. Morris USNRC Senior Resident Inspector (X24) Mr. C. Marschall USNRC Senior Resident Inspector (X24) Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625 LR-N970390 NUREG-1606 Section III.A, Definition of Change III.I, Malfunction of Equipment Important to Safety -of a Different Type ATTACHMENT PSE&G COMMENTS ON NUREG-1606 Comments The proposed guidance states that component replacement activities would be considered maintenance only if the replacement was an identical component.
'           PDR NUREG                                           Manager-1606 C                       PDR                     L i cens ing and Regulation Attachment               (1)                       -~---*~--        -*
This contradicts the guidance in Inspection Manual Part 9800 which states that maintenance activities which replace components with replacement parts procured to the same (or equivalent) purchase specification do not require a written safety evaluation to meet 10 CFR 50.59 requirements.
I1\1111 \Ill\ \Ill\ \11\IJll\\ \Ill\ \1111\11 :
As an example of a change that would not require a 10 CFR 50.59 safety evaluation, Part 9800 cites the replacement of a thermocouple with one made by a different manufacturer but which encompasses equivalent response characteristics.
The power is in your hands .
The current industry guidance states that replacing a component with an equivalent component is a maintenance activity and would not require review under 10 CFR 50.59. "Change" is defined as "an activity which affects the design, function, or method of performing the function of a system, structure or component described in the SAR." This is consistent with previous NRC guidance in Part 9800 and in NRC Inspection Procedure 37001. The new staff position would increase the number of changes requiring evaluation under 10 CFR 50.59 with no safety benefit. The proposed guidance states that, if the proposed activity could lead to a different initiator, or involves a failure mode of a different type than the types previously evaluated, then the failure results from a malfunction of a different type. The phrase ''a different initiator" is so broad that it would result Page 1 of 4
                ... -f"\{'\.i:*1
* LR-N970390 NUREG-1606 Section III.N, Licensee Practice of Deleting Information from Safety Analysis Reports PSE&G COMMENTS ON NUREG-1606 Comments in unnecessary unreviewed safety questions.
                  ~ '--*~ '*~ ! ' ., 1                                                                                                     95-2168 REV. 6/94 v~ .1v~        -
For example, consider a modification which installs an additional piece of equipment which becomes part of the reactor coolant pressure boundary.
 
Clearly the equipment should be designed in accordance with appropriate standards to withstand the required pressures and temperatures.
Document Control Desk
If
* JUL 0*7 1997 LR-N970390 C   Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. Olshan, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. D. Jaffe, Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. S. Morris USNRC Senior Resident Inspector (X24)
Mr. C. Marschall USNRC Senior Resident Inspector (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625
 
LR-N970390 ATTACHMENT PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section                                Comments III.A, Definition  The proposed guidance states that component replacement activities of Change          would be considered maintenance only if the replacement was an identical component. This contradicts the guidance in Inspection Manual Part 9800 which states that maintenance activities which replace components with replacement parts procured to the same (or equivalent) purchase specification do not require a written safety evaluation to meet 10 CFR 50.59 requirements. As an example of a change that would not require a 10 CFR 50.59 safety evaluation, Part 9800 cites the replacement of a thermocouple with one made by a different manufacturer but which encompasses equivalent response characteristics.
The current industry guidance states that replacing a component with an equivalent component is a maintenance activity and would not require review under 10 CFR 50.59. "Change" is defined as "an activity which affects the design, function, or method of performing the function of a system, structure or component described in the SAR." This is consistent with previous NRC guidance in Part 9800 and in NRC Inspection Procedure 37001.
The new staff position would increase the number of changes requiring evaluation under 10 CFR 50.59 with no safety benefit.
III.I, Malfunction  The proposed guidance states that, if the proposed activity could of Equipment        lead to a different initiator, or involves a failure mode of a Important to Safety different type than the types previously evaluated, then the failure
- of a Different    results from a malfunction of a different type.
Type The phrase ''a different initiator" is so broad that it would result Page 1 of 4
 
LR-N970390                   PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section                                Comments in unnecessary unreviewed safety questions. For example, consider a modification which installs an additional piece of equipment which becomes part of the reactor coolant pressure boundary. Clearly the equipment should be designed in accordance with appropriate standards to withstand the required pressures and temperatures. If
* the failure of this new piece of equipment is to be considered "a different initiator," the proposed guidance would require a determination that there is an unreviewed safety question, even though the consequences of its failure (small or large break LOCA) are already bounded in the accident analyses.
* the failure of this new piece of equipment is to be considered "a different initiator," the proposed guidance would require a determination that there is an unreviewed safety question, even though the consequences of its failure (small or large break LOCA) are already bounded in the accident analyses.
In practice, the proposed guidance would require prior NRC approval for many design changes that add new components or replace existing components.
In practice, the proposed guidance would require prior NRC approval for many design changes that add new components or replace existing components. In determining if the possibility of a malfunction of a different type may be created, the focus of the 10 CFR 50.59 Safety Evaluation should be on the effects of the proposed change.
In determining if the possibility of a malfunction of a different type may be created, the focus of the 10 CFR 50.59 Safety Evaluation should be on the effects of the proposed change. The proposed guidance states that licensees not remove material from safety analysis reports unless the material is changed as a direct result of a change to the facility." This conflicts with the January 1984 Part 9800 inspection guidance in which the NRC recognized that not all information contained in the SAR was used to establish the basis for the plant Operating License. The NRC stated that the intent of 10 CFR 50.59 is to limit the requirement for writing safety evaluations to facility changes, tests and experiments which could impact the safety of operations.
III.N, Licensee    The proposed guidance states that licensees ~may not remove material Practice of        from safety analysis reports unless the material is changed as a Deleting Information from Safety Analysis Reports direct result of a change to the facility."
10 CFR 50.59 should not be interpreted to prevent licensees from removing information from the FSAR that does not effect plant safety. Page 2 of 4
This conflicts with the January 1984 Part 9800 inspection guidance in which the NRC recognized that not all information contained in the SAR was used to establish the basis for the plant Operating License. The NRC stated that the intent of 10 CFR 50.59 is to limit the requirement for writing safety evaluations to facility changes, tests and experiments which could impact the safety of operations.
* LR-N970390 NUREG-1606 Section III.O, Application of 10CFR50.59 to the Resolution of Degraded and Nonconforming Conditions III.S, Definition of Reduction in Margin of Safety PSE&G COMMENTS ON NUREG-1606 Comments The proposed guidance would require a 10 CFR 50.59 safety evaluation to be performed whenever a degraded or nonconforming condition is not resolved "at the first available opportunity." A plant currently operating with a degraded condition involving a USQ would not normally be required to shutdown as long as all necessary equipment is operable.
10 CFR 50.59 should not be interpreted to prevent licensees from removing information from the FSAR that does not effect plant safety.
However, for shutdown plants, the NRC staff would not allow startup until the condition was first corrected or staff approval was received.
Page 2 of 4
 
                                                                                          \ -
LR-N970390                   PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section                               Comments III.O, Application The proposed guidance would require a 10 CFR 50.59 safety evaluation of 10CFR50.59 to    to be performed whenever a degraded or nonconforming condition is the Resolution of  not resolved "at the first available opportunity." A plant Degraded and        currently operating with a degraded condition involving a USQ would Nonconforming      not normally be required to shutdown as long as all necessary Conditions          equipment is operable. However, for shutdown plants, the NRC staff would not allow startup until the condition was first corrected or staff approval was received.
The proposed staff position is inconsistent with the proposed guidance for discovery of an inadequate Technical Specification in Section III.L. In that case, upon discovering the inadequate Technical Specification, the licensee should take appropriate action to put the plant in a safe condition (such as by imposing more conservative administrative limits), and also take action (such as requesting a license amendment) so that the Technical Specifications represent the minimum requirements.
The proposed staff position is inconsistent with the proposed guidance for discovery of an inadequate Technical Specification in Section III.L. In that case, upon discovering the inadequate Technical Specification, the licensee should take appropriate action to put the plant in a safe condition (such as by imposing more conservative administrative limits), and also take action (such as requesting a license amendment) so that the Technical Specifications represent the minimum requirements.
The existence of an unreviewed safety question does not mean that a \ -* safety issue exists, but only that NRC review is required before the
The existence of an unreviewed safety question does not mean that a safety issue exists, but only that NRC review is required before the
* change is implemented.
* change is implemented. Upon discovery of a degraded or nonconforming condition which involves an unreviewed safety question, licensees should make timely application for NRC review and approval. Provided all necessary equipment is operable, a degraded condition involving a potential USQ would not require a plant to remain shutdown.
Upon discovery of a degraded or nonconforming condition which involves an unreviewed safety question, licensees should make timely application for NRC review and approval.
III.S, Definition  The proposed requirement to ref er to plant-specific SAR values as of Reduction in    acceptance limits represents a new requirement for two reasons.
Provided all necessary equipment is operable, a degraded condition involving a potential USQ would not require a plant to remain shutdown.
Margin of Safety First, as the NRC noted in the April 1996 Part 9900 inspection Page 3 of 4
The proposed requirement to ref er to plant-specific SAR values as acceptance limits represents a new requirement for two reasons. First, as the NRC noted in the April 1996 Part 9900 inspection Page 3 of 4 LR-N970390 NUREG-1606 Section III.V, Consideration of Compensating Effects PSE&G COMMENTS ON NUREG-1606 Comments guidance, industry guidance is currently broader than the rule regarding where a licensee must look to find a margin of safety in that NSAC-125 recommends looking beyond the Technical Specification (TS) Bases. In stating that 10 CFR 50.59 requires evaluation of margins of safety not specifically described in the TS Bases, the
 
LR-N970390                     PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section                                Comments guidance, industry guidance is currently broader than the rule regarding where a licensee must look to find a margin of safety in that NSAC-125 recommends looking beyond the Technical Specification (TS) Bases. In stating that 10 CFR 50.59 requires evaluation of margins of safety not specifically described in the TS Bases, the
* proposed guidance imposes new requirements for the conduct of 10 CFR 50.59 Safety Evaluations.
* proposed guidance imposes new requirements for the conduct of 10 CFR 50.59 Safety Evaluations.
Second, the proposed guidance would require a proposed change to be identified as an unreviewed safety question if it involved a potentially non-conservative change in a value in the SAR (as referenced in the Technical Specification Bases) . The acceptance limit is not typically the value reported in the SAR. Changes in SAR reported values (e.g., ECCS pump available net positive suction head) do not involve a reduction in the margin of safety unless they exceed the acceptance limit reviewed and approved by the NRC. The proposed guidance states that the "effect of any change must be evaluated against each of the USQ criteria separately  
Second, the proposed guidance would require a proposed change to be identified as an unreviewed safety question if it involved a potentially non-conservative change in a value in the SAR (as referenced in the Technical Specification Bases) . The acceptance limit is not typically the value reported in the SAR. Changes in SAR reported values (e.g., ECCS pump available net positive suction head) do not involve a reduction in the margin of safety unless they exceed the acceptance limit reviewed and approved by the NRC.
-that is, an increase in probability cannot be 'compensated' by additional mitigation capability." This conflicts with previous NRC guidance on the use of compensatory actions to offset increases in probability or consequences.
III.V,              The proposed guidance states that the "effect of any change must be Consideration of    evaluated against each of the USQ criteria separately - that is, an Compensating Effects increase in probability cannot be 'compensated' by additional mitigation capability." This conflicts with previous NRC guidance on the use of compensatory actions to offset increases in probability or consequences. The April 1996 Part 9900 inspection guidance states that it is acceptable to use compensating effects (such as changes in administrative controls) to offset uncertainties and increases in probability or consequences. A requirement to evaluate compensatory measures separately from the proposed change would, in many cases, prevent the use of administrative controls as a compensating effect.
The April 1996 Part 9900 inspection guidance states that it is acceptable to use compensating effects (such as changes in administrative controls) to offset uncertainties and increases in probability or consequences.
Page 4 of 4}}
A requirement to evaluate compensatory measures separately from the proposed change would, in many cases, prevent the use of administrative controls as a compensating effect. Page 4 of 4 *}}

Latest revision as of 05:12, 3 February 2020

Comment Opposing NUREG-1606, Proposed Regulatory Guidance Re Implementation of 10CFR50.59 (Changes,Tests or Experiments). Util Endorses Comments Submitted by Nuclear Energy Inst
ML18102B436
Person / Time
Site: Salem, Hope Creek  PSEG icon.png
Issue date: 07/07/1997
From: Dawn Powell
Public Service Enterprise Group
To:
NRC OFFICE OF ADMINISTRATION (ADM)
References
FRN-62FR24997, RTR-NUREG-1606 62FR24997-00029, 62FR24997-29, LR-N970390, NUDOCS 9707150120
Download: ML18102B436 (6)
v • d • e


Text

  • iJSt'lf
  • 6_2;~ -2f/2Y7 0 Pst&G ~

E. ~11£'/l/Jc, S'-//e~y

('. /7?/m:Y)

, /Yo-y ~ /777 Public Service Electric and Gas Company

Nuclear Business Unit JUL 0 7 1997 LR-N970390 Chief, Rules Review and Directives Branch Division of Administrative Services Off ice of Administration :0  ::0 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

~ ~c:::~ ::D v.; tt{i cs;:o .- ("')

(/)0 SALEM GENERATING STATION UNITS 1 AND 2 .LSJ HOPE CREEK GENERATING STATION :IJ* ~

FACILITY OPERATING LICENSES. DPR-70 I DPR-75 AND NPF-57 Oco-~*J SC ""'""'

rn

'}~ -..o*

DOCKET NOS. 50-272, 50-311 AND 50-354 Z *** CJ COMMENTS ON NUREG-1606, "PROPOSED REGULATORY GUIDANCE REubf.ED~*o IMPLEMENTATION OF 10 CFR 50. 59 (CHANGES, TESTS, OR EXPERIMENTS)"

Gentlemen:

Public Service Electric and Gas Co. (PSE&G) provides the attached comments in response to the Federal Register notice dated May 7, 1997. I Our comments are directed towards those sections of the NUREG which we believe to be new requirements. Their effect would be to greatly increase the number of proposed changes requiring NRC review and approval with little or no safety benefit.

In addition to the detailed comments attached to this letter, PSE&G also endorses the comments submitted by the Nuclear Energy Institute and by the Licensing and Design Bases Clearinghouse.

PSE&G appreciates the opportunity to provide comments on the draft document. Should there be any questions concerning this submittal, please do not hesitate to contact us.

Sincerely,

(,.-------- -9707150120 970707 David R. Powell

' PDR NUREG Manager-1606 C PDR L i cens ing and Regulation Attachment (1) -~---*~-- -*

I1\1111 \Ill\ \Ill\ \11\IJll\\ \Ill\ \1111\11 :

The power is in your hands .

... -f"\{'\.i:*1

~ '--*~ '*~ ! ' ., 1 95-2168 REV. 6/94 v~ .1v~ -

Document Control Desk

  • JUL 0*7 1997 LR-N970390 C Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. L. Olshan, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. D. Jaffe, Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. S. Morris USNRC Senior Resident Inspector (X24)

Mr. C. Marschall USNRC Senior Resident Inspector (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625

LR-N970390 ATTACHMENT PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section Comments III.A, Definition The proposed guidance states that component replacement activities of Change would be considered maintenance only if the replacement was an identical component. This contradicts the guidance in Inspection Manual Part 9800 which states that maintenance activities which replace components with replacement parts procured to the same (or equivalent) purchase specification do not require a written safety evaluation to meet 10 CFR 50.59 requirements. As an example of a change that would not require a 10 CFR 50.59 safety evaluation, Part 9800 cites the replacement of a thermocouple with one made by a different manufacturer but which encompasses equivalent response characteristics.

The current industry guidance states that replacing a component with an equivalent component is a maintenance activity and would not require review under 10 CFR 50.59. "Change" is defined as "an activity which affects the design, function, or method of performing the function of a system, structure or component described in the SAR." This is consistent with previous NRC guidance in Part 9800 and in NRC Inspection Procedure 37001.

The new staff position would increase the number of changes requiring evaluation under 10 CFR 50.59 with no safety benefit.

III.I, Malfunction The proposed guidance states that, if the proposed activity could of Equipment lead to a different initiator, or involves a failure mode of a Important to Safety different type than the types previously evaluated, then the failure

- of a Different results from a malfunction of a different type.

Type The phrase a different initiator" is so broad that it would result Page 1 of 4

LR-N970390 PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section Comments in unnecessary unreviewed safety questions. For example, consider a modification which installs an additional piece of equipment which becomes part of the reactor coolant pressure boundary. Clearly the equipment should be designed in accordance with appropriate standards to withstand the required pressures and temperatures. If

  • the failure of this new piece of equipment is to be considered "a different initiator," the proposed guidance would require a determination that there is an unreviewed safety question, even though the consequences of its failure (small or large break LOCA) are already bounded in the accident analyses.

In practice, the proposed guidance would require prior NRC approval for many design changes that add new components or replace existing components. In determining if the possibility of a malfunction of a different type may be created, the focus of the 10 CFR 50.59 Safety Evaluation should be on the effects of the proposed change.

III.N, Licensee The proposed guidance states that licensees ~may not remove material Practice of from safety analysis reports unless the material is changed as a Deleting Information from Safety Analysis Reports direct result of a change to the facility."

This conflicts with the January 1984 Part 9800 inspection guidance in which the NRC recognized that not all information contained in the SAR was used to establish the basis for the plant Operating License. The NRC stated that the intent of 10 CFR 50.59 is to limit the requirement for writing safety evaluations to facility changes, tests and experiments which could impact the safety of operations.

10 CFR 50.59 should not be interpreted to prevent licensees from removing information from the FSAR that does not effect plant safety.

Page 2 of 4

\ -

LR-N970390 PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section Comments III.O, Application The proposed guidance would require a 10 CFR 50.59 safety evaluation of 10CFR50.59 to to be performed whenever a degraded or nonconforming condition is the Resolution of not resolved "at the first available opportunity." A plant Degraded and currently operating with a degraded condition involving a USQ would Nonconforming not normally be required to shutdown as long as all necessary Conditions equipment is operable. However, for shutdown plants, the NRC staff would not allow startup until the condition was first corrected or staff approval was received.

The proposed staff position is inconsistent with the proposed guidance for discovery of an inadequate Technical Specification in Section III.L. In that case, upon discovering the inadequate Technical Specification, the licensee should take appropriate action to put the plant in a safe condition (such as by imposing more conservative administrative limits), and also take action (such as requesting a license amendment) so that the Technical Specifications represent the minimum requirements.

The existence of an unreviewed safety question does not mean that a safety issue exists, but only that NRC review is required before the

  • change is implemented. Upon discovery of a degraded or nonconforming condition which involves an unreviewed safety question, licensees should make timely application for NRC review and approval. Provided all necessary equipment is operable, a degraded condition involving a potential USQ would not require a plant to remain shutdown.

III.S, Definition The proposed requirement to ref er to plant-specific SAR values as of Reduction in acceptance limits represents a new requirement for two reasons.

Margin of Safety First, as the NRC noted in the April 1996 Part 9900 inspection Page 3 of 4

LR-N970390 PSE&G COMMENTS ON NUREG-1606 NUREG-1606 Section Comments guidance, industry guidance is currently broader than the rule regarding where a licensee must look to find a margin of safety in that NSAC-125 recommends looking beyond the Technical Specification (TS) Bases. In stating that 10 CFR 50.59 requires evaluation of margins of safety not specifically described in the TS Bases, the

  • proposed guidance imposes new requirements for the conduct of 10 CFR 50.59 Safety Evaluations.

Second, the proposed guidance would require a proposed change to be identified as an unreviewed safety question if it involved a potentially non-conservative change in a value in the SAR (as referenced in the Technical Specification Bases) . The acceptance limit is not typically the value reported in the SAR. Changes in SAR reported values (e.g., ECCS pump available net positive suction head) do not involve a reduction in the margin of safety unless they exceed the acceptance limit reviewed and approved by the NRC.

III.V, The proposed guidance states that the "effect of any change must be Consideration of evaluated against each of the USQ criteria separately - that is, an Compensating Effects increase in probability cannot be 'compensated' by additional mitigation capability." This conflicts with previous NRC guidance on the use of compensatory actions to offset increases in probability or consequences. The April 1996 Part 9900 inspection guidance states that it is acceptable to use compensating effects (such as changes in administrative controls) to offset uncertainties and increases in probability or consequences. A requirement to evaluate compensatory measures separately from the proposed change would, in many cases, prevent the use of administrative controls as a compensating effect.

Page 4 of 4

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